ML20008D767

Chapter 5 to Midland 1 & 2 PSAR, Reactor Bldg & Structures. Includes Revisions 1-36

ML20008D767
Person / Time
Site:	Midland
Issue date:	01/13/1969
From:	CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
Shared Package
ML20008D768	List: ML20008D767 ML20008D769
References
NUDOCS 8007300654
Download: ML20008D767 (83)

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' (v%-1 TA3LE OF CON.T TS V

section Page 5 REACTOR BUILDING A'O SdULTJEES 5-1 +

51 EEACTOR BUILDI'D 5-1 5 1.1 ETE'ATJRAL DESIGN 5-1 5 1.1.1 General Da cription of Reactor Building 5-1 5.1.1.2 Easis for Design Lcads 5-3 5 1.1 3 construction Materials 5-6 51.1.4 Reacter Building Desien Criteria 5-15 51.15 structural Design Analysis 5-29 5 1.2 DESI. AID CONSIRUCTION OF PE:2TRATIONS 5-38 5 1.M.1 Types of Penetrations 5-38 5 1.3.2 Design of Penetrations 5-ho 5 1.2 3 Installation cf Penetrations 5-h2a

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513 coNSIRUCTION PEACTICES AID QUALITY ASSURAIDE 5-42a 5.131 Applictble construction codes 5-42a 5132 Quality Control Procram 5-k3 5.1 3 3 construction 2.nteri's.ls Inspection and Installatic 5-L3 5 1 3.h Specific Construction Tcpics 5-k7 5 1.k EEACTOR BUILDIIU INSP5$" ION, TESTI?U, AI*D SURVEILLAN~E 5-h9 5 1.k.1 Tests to Insure Liner Integrity 5 h9 5.1.h.2 Strength Test

5-32 5 1.4.3 In-service Tenden s m eillance Prcera: 5-52

• i 5 1.k.L _ Leakage Monitoring Syste: 5-33 515 IsoIATION system  : 5-55 -

u' 5151 Desien Bases '

5-55 5152 syste: Design 5-55 >

f~5 t Jl 5153 Penetratien Pressurl:ation syste: 5-56 v -

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4 A- TABIE OF COICE'ZS (Contd)

Section Page 5 1.6 DESIGN EVAIDATION 5-62 52 AUXILIA.P' 3'JIDEC 5-62 5 2.1 DESIGN BASES 5-62 5 2.2 DESIGN CRuulA AID GENERAL DESCRIPTION 5-63 l 5.2 3 ITJCLEAR FUEL STORAGE CONSEERATIONS 5-64 53 OIEER PLA!E STRU3TPSS 5-64 i

, 531 TUREINE 3JILDEG 5-6L 532 SERVICE WATER PU!G STRUCTURES 5-6L 533 DIESEL GE::IRATOR SUIIDIIn 5-65 53.4 ADMINISTPJCION AID SERVICE J3'ILDIIUS 5-65 5.L RADIATION PROTECTION 5-65 4

t ,/ 5.k.1 RADIATION ZONDG AID ACCESS COICROL 5-65

[ 5.h.2 RADIATION SmnHG 5-66 5 4.2.1 Design sases 5-66 5.L.2.2 ceneral Descripticas and Evaluations 5-67 5.k.3 it::EudCES 5-70 5.L.3 1 List cf Publications Used Extensively in the Scielcing Desisn 5-70 5.h.3.2 List of Cc=puter codes used for the Shielding Desis: 5-70 i

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Figure No. Title 5-1 Reactor Building Typical Details 5-2 Reactor Building Penetration Details 5-2A Types of Flued Head Fittings for Containment Piping i Penetrations, Units 1 & 2 26

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5-2B Typical Forged Flued Head Type Pipe Penetration Assembly 5-3 Reactnr Building Details of Equipnent Hatch 5-4 Design Thermal Gradient Across Reactor Building Wall 5-5 Long-Term Effect Creep and Shrinkage 1 5-6 Liner Plate Loading Conditions 5-7 ,

Angle Anchor Welds Test Results f 5-8 Reactor Building Isolation Valve Arrgt i

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l' LIST OF APPENDICES ,

, Number Title j SA- Design Bases for Structures, Syster.s and Equipment SB Reactor Building Pressure Tests j SC Mechanical Splicing of Rein'orcing Bar i

SD Load Factors and Load Combinations i

SE Yield Reduction Factors

. 5F Reactor Building Structural Analysis t

4 SG Quality Control Procedure for Field Welding of Liner Plate i

! 5H Reactor Building Load Response Test j i

, 25 SI Design Features.for High Energy Pipe Failure Outside the l Rea ctor - Building 1- .

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IISI 0F FIGUFIS l (At Reer Of Appendices) i I Figure No. Title 4

l 5A-1 Design Response Spectru " Design Earthquake" 1

5A-2 Design Response Spectru: "yaxi=um Earthquake" f

I Sc-1 cadweld series 3 connection 5H-1 Reactor Building Test Instrumentation s

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( 5 REACTOR BUIIENG AND STRUCTUEES

~5 1 REACTOR 3UIlDI?G 5 1.1 STRUCTURAL DESIGN 5 1.1.1 General Description of Peactor Buildins The reactor building is a fully continuous reinforced concrete structure in the shape of a right cylinder with a shallow dc=ed roof and a flat foundation slab. The cylindrical portion is prestressed by a post-tensioning syste= con-sisting of hori: ental and vertical tendens. The do=e has a three-way post-i tensioning syste=. Hoop tendens are placed in 3 - 240 degree syste=s using

I three buttresses as anchorages, with the tendons staggered so that two thirds of the tendons at each buttress terminate at that buttress. The foundation slab is conventionally reinforced with high-stren6th reinforcing steel. A continuous access gallery is previded beneath the base slab for installation and inspection of vertical tendons. The inside face of the concrete shell is steel lined to insure a high degree of leak tightness. The base liner with leak chases at joints is installed on tcp of the structural slab and is covered with concrete. The structure provides biological shielding for both nornal and accident situations.

The reactor building co=pletely encloses the entire reactor and reactor coolant system and insures that an acceptable upper limit for leakage of g radioactive materials to the environ =ent would not be exceeded even if gross

") failure of the reacter coolant syste= vere to occur. The apprcxi= ate di=en-sions of the reactor building are: inside dia=eter, 116 feet; inside height, 193 feet; vertical vall thickness, 3-1/2 feet; and the dome thickness, 3 feet.

The building encloses the pressuriced water reactor, stea: generators, reactor coolant loops and portions of the auxiliary and engineered safeguards syste=s.

The internal net free v0.w=e is 1,670,000 cubic feet.

Full advantage is being taken in the design of this reacter cuilding of the experience gained in the review of similar designs with the AEC for the Florida Power and Light Cc=pany's Turkey Point Plant, Wisconsin Michigan Power Cc=pany's Point Beach Plant, Duke Power Conpany's Oconee Nuclear Station, Arkansas Power and Light Conpany's Russellville Plant, the Sacra-

= ento Municipal Utilities District Rancho Seco Plant and the Palisades Plant, as well as reactor building designs by others which =eet the sc=e functional requirements.

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( nepresentative details cf the construction thst vill be used are shown in i Figures 5-1, 5-2 and 5-3.

Table 5-1 cc= pares principal differences ' cf the Midland react ,r tuilding with reactor buildings cf other plants.

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Table 5-1 Comparinons With Other Henctor Buildings Consumers Power Co Consumern Power Co Arkannan Power & Light Co Midland Plant, Palinnden Plant Runnellville 'lant Diameter 116'-0" 116'-0" 116'-0" lD;olgnPrennure 67 Paig 55 Paig 59 Paig Wall Thicknena 3'-6" 3'-6" 3'-9" Dome ThIckneno 3'-0" 3'-0" 3'-3" Level of Prentrenn 1.2P-1.5P(3) 1.5P 1.5P lWiresperTendon 170 (Maximum) 90 llM Gize of Wire 1/h" 1/h" 1/h" Buttressen 3 6 3 y Gacramento Municipal su Ut.111ty Dintrict, Florida Power & Light Co Wisconnin Michigan Power Co Hancho Seco Turkey Point Point Bench Diameter 130'-0" 116'-0" 105'-0" Design Pressure 59 Poig 59 Paig 60 Poig g 3'-9" 3'-6"

.+ Wall Thicknesn 3'-9"

[ Dome Thicknean 3'-0" 3'-3" 3'-0" (q Level of Prentress 1.2"-1.5P(1) 1.5P 1.5P Wires per Tendon 55 Strando 90 90 Size of Wire 1/2" (T Wire) 1/h" 1/h" Buttressen 3 6 6 ls n

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M Cracking of the concrete in allowed under the effective lond condition with the membrnne forcen entirely 08- resist.ed by prent.renalng nt. eel and bonded reinforcing steel.

5.1.1.2 Basis for Desien Loads O V The reactor building is designed for all credible conditions of loading, in-cluding normal loads, loads during loss-of-coolant accident, test loads, and loads due to adverse environ = ental conditions. The following loadings are considered:

a. The Loadings Caused by the Pressure and Temperature Trt.nsients of the Loss-of-Coolant Accident (LOCA)

b. Structure Dead Load

c. Live Loads

d. Earthquake Load

e. Wind Force and Tornado Loads

f. Uplift Due to Buoyant Forces

g. External Pressure Load The two critical loading conditions are those caused by the loss-of-coolant accident resulting from failure of the reactor coolant system and those caused by an earthquake.

5.1.1.2.1 Loss-of-Coolant Accident Load p) t V

The minimum design pressure and temperature of the reactor building are equal to the peak pressure and temperature occurring as the result of any rupture of the reactor coolart system up to and including the severance of a reactor coolant pipe.

The supports for the reactor coolant system are designed to withstend + %

blevdown forces associated with the sudden severance of the reactor coolat.t piping so that the coincidental rupture of the steam and feedvater systems is not considered credible.

As noted in the Section lb Accident Analysis, the peak accident pressure in the reactor building is calculated to be 60.0 psig for the 5.0 sq ft break LOCA, according to the CONTIMFT code used by the AEC. Adding the 10 percent margin required by the AEC yields 66.0 psig, and 1.0 psi is incorporated for additional margin to arrive at a reactor building design pressure of 67 psig.

The te=perature gradient through the vall during the loss-of-coolant accident is shown in Figure 5 h. The variation of te=perature with time and the expan-sion of the liner plate are considered in designing for the thermal stresses as sociated with the LOCA.

5.1.1.2.2 Structure Dead Load Dead load consists of the weight of the concrete vall, dome, base slab, and any internal concrete. Weights used for dead load calculations are as follows:

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\ 2 V- a. Concrete lh8 lb/ft"

b. Steel Reinforcing kB9 lb/ft 3, using nominal cross-sectional areas of reinforcing as defined in ASTM for bar sites and nominal cross-sectional area,s of prestressing

e. Steel plate E89 lb/ft 3, using nominal cross-sectional area 5 1.1.2 3 Live Icaas Live loaca, which consist cf all loads except dead, accident, seismic, flood and vind f.nclude sn v loads on the partially domed roof cf the reactor building. The roof lead is 40 pounds per horizontal square foot.

Equipment loads are those specified on the drawings supplied by the manu-

. facturers of the various pieces of equipment.

Live leads are assu=ed for the design of internal slabs consistent with the intended use of the slabs.

5 1.1.2.4 Earthquake Loads hbgnitudes cf ground accelerations due to " Design Earthquake" and "thxi==

v z Earthquake" are- given in Appendix 5-A together with design response spectr=

curves.

A vertical ec=ponent two-thirds of the =agnitud:e of the horizontal e:eponent is applied in the 1 cad equations simultaneously. A dynatic analysis is used to arrive at equivalent static loads fer design.

5 1.1.2 5 Wind and Tornado Loads Wind loading is 65 =ph basic wind at 30 feet above grade, based en Figure. l (b)

SC ASCE paper 3269, " Wind Forces on Structures." This vind load is considered for the reactor building design.

Tnere are few reliable *easure=ents of the pressure drop associated with a terr. ado' funnel. Tne reatest drop reported was equivalent to a bursting presste e of approxP ately 3 psi. This measurenent, however, is highly questionable av .s not regarded as: authoritative. Tne greatest Prestare dreps which have been reliably neasured are On the order cf 15 psi.

Tne maxi =u design bursting pressure difference is 3 psi. It.is 100 percent greater than the-greatest pressure ever reliably measured. Tnis value is thought to be very conservative.

Because cf the-ec7 1exity of the airflow in a tornado, it has-not been

-possible to calculate the velocity or trajectory of =1ssiles that would truly

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A I represent tornado conditions. Ftr design purposes, it is conservatively assu=ed that objects of low cross-sectional density such as boards, =etal siding, and similar iters may be picked up and carried at the maxi =ur vind velocity of 300 =ph.

The behavier of heavier, irregularly shaped objects such as an automobile caught in a tornado is not predictable. The design value of 50 mph for a 4000 pound auto =obile lifted 25 feet in the air is felt to be a conservative representation of what vould happen in a 300 =ph vind as the autocobile was lifted, tu= bled along the ground, and ejected ' " e tornado funnel by centrifugal force. These missile velocities are conservative when compared with observed phenomena of previous tornadoes.

The structure is analyned for tornado loading (not coincident with accident or earthquake) on the following basis :

a. Differential bursting pressure between the inside and outside of the reactor building is assu=ed to be 3 pounds per square inch positive pressure.

b. Lateral force on the reactor building is assu=ed as the force caused by a tornado funnel naving a =axi=u peripheral tangential velocity of 300 mph and a forward progression of 60 mph. These velocity co=-

ponents are conservatively applied as a 300 =ph vind over the entire

, surface of the structure for each reactor building and are additive for a 360 =ph wind over the entire surface for other class 1 strue-

' tures. The applicable portions of vind design methods described in ASCE Paper 3269 vill be used, particularly for shape factors. The N~ provisions for gust factors and variations of vind velocity with neight do not apply.

c. Tornado-driven =issiles equivalent to an airborne k inch by 12 inch by 12 foot plank traveling end-or at 300 =ph, or a k000 pound auto-

=obile flying through the air at 50 =ph and at not = ore than 25 feet above the ground are assumed.

A discussion of the probability of tornado occurrence is presented in Appendix 2-A. .

Except for local crushing at the missile impact area, the allevable stresses to resist the effects of tornadoes are 90 percent of the yield of the rein-forcing steel and 75 fercent of the ultimate strength of the concrete.

15 1.1.2.6 Uplift Due to Buoyant Fcrces Uplift forces which are created by the displace =ent of groundwater by the structure are accounted fer in the design of the structures.

5 1.1.2 7 External Pressure Load External pressure loading with a differential of 2-1/2 pounds per square-inch fro outside-to inside is considered.

'The external design pressure is equivalent to having a barc=etric pressure f' rise to 3h inches of mercury after the resctor building was sealed at 29 inches 3-+).

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I of mercury. Therefore, operation of purge valves will not be required due l to barometric changes during normal operation.

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The external design pressure is also adequate to per=dt the reactor building to be cooled to 50 F, which is a reasonable winter internal temperature during shutdown, from an initial maximum operating condition of 120 F.

Therefore, operation of purge valves will not be necessary during this shutdown condition. Vacuu breakers are not required.

5.1.1. 3 Construction Materials Basically These are: four materials are used for the ft.adation and the reactor building.

a. Concrete

b. Reinforcing Steel

c. Steel Prestressing Tendons

d. Reactor Building Liner Detailed specifications and working drawings for these ma:erials and their installation are of such scope as to assure that the quality of work is com-mensurate with the necessary integrity of the reactor building.

Basic specifications for these materials follow:

/'7q 5.1.1.3.1 Concrete

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' All concrete work is in accora.nce with ACI 318-63, " Building Code Require-ments for Reinforced Conerste" nd ACI-301, " Specifications for Structural Concrete '?or Buildings." Admixt *es are added to improve the quality and workability of the plastic concre e during placement. Maximu= practical size 2

aggregate, water-reducing additivaa, and a low slump of two or three inches are used to minimize shrinkage and creep. Aggregates conform to " Standard Specifications for Concretr. Aggregate" ASTM Designation C-33-71a, as modified 27 by paragraph 8.3 of the tentative revision to ASTM C-33-71a, published with ASTM C-33-71a in the 1973 annual ASTM Standards, Part 10.

Acceptability of aggregates is based on the following ASTM tests. These teste 'are performed by a qualified commercial testing laboratory, and results

' submitted (1) to the owner, and (2) the Quality Assurance Engineer of the ConstIJCtor.

Test ASTM L. A. Rattler C-131 Clay Lumps Natural Aggregate C-142 Material Finer No. 200 Sieve C-ll7 Mortar-Making Properties C-87 Organic Impurities C-40 '

- Potential Reactivity (Chemical) C-289 Potential Reactivity (Mortar Bar) C-227 (If necessary af ter perform-(T ing C-289)

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Test ASTM l Sieve Analysis C-136  :

Soundness C-88 Specific Cravity and Absorption C-127 Specific Gravity and Absorption C-128 Petrographic C-295 Cenent is-Type II low alkali cement as specified in " Standard Specifications for Portland Cement" ASTM Designation C-150 and is tested to comply with ASTM C-ll4 Water for nixing concrete is clean and free fro: any deleterious amounts of acid, alkali, salts, oil, sediment, or organic =atter.

A water-reducing agent is e= ployed to reduce shrinkage and creep of concrete.

27l Admixtures containing chlorides shall conform to applicable ASTM standards. '

h.. following types of agents are tested with the concrete materials selected for the reactor building:

5 P zzolith No. 82 and Pozzolith 100R (Mfd by Masters Builders Co=pany)

Daratard .fd by Grace Construction Mf g)

Flastimes. (Mfd by Sika Chemical Co=pany)

Placewell LS (Mfd bf Union Carbide)

The agent selected is the one p.oviding a small shrinkage as deter =ined by 5 ASTM C-494, Type "D," " Specifications for Chemical Admixtures for Concrete,"

togethe: with workability.

I- ' Ccncrete mixes are designed in accordance with ACI 613 using materials qualified and accepted for this work. Only =ixes meeting the design require-ments specified for reactor building concrete are used. Trial mixes are tested in accordance with applicable ASTM Codes as indicated beloe*:

Test ASTM Making and Curing Cylinder in C-192 Laborato ry Air Content C-231 Slu=p C-143 27l DELETED Compressive Strength Tests C-39 4

Six cylinders are cast f rom each design mix for two tests on each of the following days: 3, 7, 28 and 90 days. Three-day test will only be made on occasion to correlate three-day strength. The concrete has a design con-

, pressive strength of e000 psi at 90 days for the reactor building wall and 47 dome and 4000 psi at 90 days for the reactor building base slab.

Test cylinders are cast from the mix proportions selected for construction and the following concrete properties are determined:

Uniaxial creep

,_ Modulus of elasticity and Poisson's ratio 1

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Autogeneous shrinkage Thermal diffusivity Thermal coefficient of expansion

-((w) v Co=pressive strength Aggregate testing is carried out as follows:

a. Sand Sample for Gradation (ASTM C-33 Fine Agg) (2 tests per shift)

b. Organic Test on Sand (ASTM C-40) (1 test per shift)

c. 3/4 Inch Sample for Gradation (ASTM C-33 Size No. 67) (1 test per shift)

d. 1-1/2 Inch Sample for Gradation (ASTM C-33 Size No. 4) (1 test per shift)

e. Check for Proportion of Flat and Elongated Particles (1 test per week)

Concrett samples are taken from the mix every 100 cubic yards according 3 to ASTM C-172, " Standard Mathod of Sampling Fresh Concrete". From these samples, cylinders for compression testing are made. They are stripped within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> af ter casting and marked and stored in the curing room.

These cylinders are =ade in accordance with ASTM C-31, " Standard Method for Making and Curing Concrete Compression and Flexure Test Specimens in the Field".

.(\-A') Slump is taken for every 35 yards produced for Class 1 structures and air content, and temperature measurements are taken when cylinders are cast.

Slump tests are performed in accordance with ASTM C-143, " Standard Method of Test for Slump of Portland Cement Concrete". Air content tests are per-formed in accordance with ASTM C-231, " Standard Method of Test for Air Con-tent of Freshly Mixed Concrete by the Pressure Method". Compressive strength tests are made in accordance with ASTM C-39, " Standard Method of Test for Compressive Strength cf Molded Concrete Cylinders".

Evaluation of compression tests is in accordance with ACI 214-65.

5.1.1.3.2 Reinforcing _ S teel 2 Reinforcing steel in the base slab of the reactor building and around penetra-tions in the cylindrical wall is deformed billet steel bars conforming to 261 either ASTM Designation A-615-68 or A-615-72, Grade 60. This steel has a

,l minimum yield strength of 60,000 psi, a minimum tensile strength of 90,000 psi,

'I and a minimum elongation of 7 percent in an 8 inch specimen. Deformed billet steel bars conforming to either ASTM A-615-68 or A-615-72, Grade 60 are used in the cylindrical wall and the domed roof to control shrinkage and tensile cracks.

Mill test results are obtained from the reinforcing steel supplier for cach heat of steel to show proof that the reinforcing steel has the specified compo-sition, strength, and ductility. Testing of reinforcing steel shall be in accord-26 a e with ASTM A-615. The supplier's tensile tests may utilize the option given f-- in the ASIM A-615 Section_ entitled " Test Speciments". User's tests shall be s(x-s)' made on full size diameter bar utilizing the 8 inch gange length for each 50 tons, or fraction thereof, of -

03F,2 Amendment No. 26 5-8 4/ 74

26 reinforcement from each heat. Splices in reinforcing bar Sizes No. 11

[s'~ N) and smaller are lapped in accordance with either ACI 318-63 or ACI 318-71, and for bars larger than No.11, CADWELD splices are used in accordance with Appendix 5-0.

Welding of reinfc rcing steel, although not intended, if required, is per-formed by qualifj ed welders in accordance with AWS 912.1, " Recommended Practice for Welding Reinforcing Steel, Metal Inserts, and Connections in Reinforced Concrate Construction".

5.1.1. 3. 3 Steel Prestressing Tendons The following describes the tendon p erformance requirements for this plant 5 land the characteristics of the prestressing system.

a. Tendon Sizes One performance requirement is that tendons, v! . ien integrity characteristics, be available in a range of str gth capabilities.

With this requirement satisfied, the basic tendon for the design can be chosen as the one with the largest strength compatible with the containment being designed and, should the need arise, smaller ones could be used if needed in special limited space situations.

The tendou compatibility with the structure size has been evaluated.

It was found that the tendon strength is compatible with this s -5 structure, and the 170 wire strength capabilities are preferable

for this structure. The following items will remain unchanged

(1) The corrosion protection system.

(2) The structural test.

(3) The number of tendons subjected to the surveillance program.

(4) The procedure for development and demonstration of field equipment.

(5) The construction procedure of placing and stressing of tendons.

(6) - The size of the tendon access gallery appears to be controlled by buttonheading space requirements. Any increase in size to accommodate the jack will be nominal, probably not to exceed one foot in width or height.

In establishing acceptable tendons, the following was considered:

(1) The nominal vertical center to center spacing of 90 wire cir-cumferential tendons is less than 10 inches, while for the 170 5 wire tendon .the spacing is about 20 inches. Both nominal spacings are less than one-half of the 42 inch cylinder wall thickness and result in reasonably uniform prestressing force

,-~s application to the cylinder. However . the st

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20 inch spacing results in a greater ease fer concrete place-

=ent and lesser embed ent 00ngestien near penetrations.

Further, the redundancy Of the wires is not reduced by the use of the 170 vire tendons as conpared to the 90 wire tendens since the total number cf wires renains the sane. As an illustration of the redundancy of tendens, three adjacent tendens could be re=cred without significantly af-fecting the structural integrity because of the stiffness of the shell between the re=aining tendens.

(2) The average bearing pressure against the tenden cheaths for this structure is one-fifth cr less Of that which occurred during tests of the 170 vire tendon with 160 degree curves with 10 foot radius. Even at the penetrations, the s=allest radius of curvature for this contain=ent, the bearing pressure under the sheath is smaller than that for the 10 foot radius tension test of the 170 vire tenden. Regardless of size of the individual tendon, the pattern of =e=brane stresses, bend-ing stresses, and radial co=pression er tension stresses at the various penetrations are not expected to be caterially affected due to the small ratic. of the tendon spacing to vall thickness and the shell stiffness as cited above. Such aspects vill be considered during the detailed design of the structure.

(3) Prestressing tendon syste= develop =ent for large tendons has

(~'N been done pri=arily in suppcrt of prestressed corerete reactor

-(,) vessel verk in France, England and the United States. For exa ple, the Dungeness 3 nuclear power plant in England is using a syste= with about 160 vires, 7 == (0.276 in.) in dia=-

eter, which corresponds in force capacity te a tendon with 130 or =cre 0.25 inch diameter wires. The jack fer this size tenden has been extensively tested as a part of the syster develop =ent. The 170 re tenden has been deve10 ped for the prestressed concrete ivattor vessel for the Fort Saint vrain plant and stressing equipment has provided 1000 cycles of 1 cad applicaricns, =0re than sufficient to stress all tendens on this reacter building, in one test series alone.

b. End Anchers, Eearing Plates and Prestressing Steel The basic perfer=ance requirements for the end anchers are stated qualitatively by the Seis=10 C0==1ttee of the Prestressed Cenerete Institute and published in their Journal of June 1966 as fellows:

"All anchors cf unb0nded tendons should develop at least ICO percent of the guaranteed c'.ti= ate strength cf the tendon. The anchcrage gripping shouli function in such a way that no har ful notching would occur on the tendon.

Any such anchorage syste= used in 2arthquake areas cust be capable of =aintaining the prestressing force under sus-tained and fluctuating load and the e 'e t Of shock. Anchors f'"3 i i

should also possess adequate reserve strength to withstand

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any everstress ta which they may be subjected :iuring the =cs:

, (~ severe prebable ea-thquake. Particular care shoul'. be directed N to accurate positiening and align =ent of end anchers."

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. ~. ~. s .= .

.d ..^,~..-' e . '. c .. v'. . e .= . .= .' ~y . % . .. %m required fer any tenden syste: used for this structure.

It is a performance require:ent that the end anchers used develop 100 perent of the guarantee' ~' *- ulti=ste strength cf the prestressing steel withcut permanent deformatic: which renders the end ancher u= sable er results in significant slip of the prestressirg steel relative to the end ancher. The test results of available systers give evidence that those end anchers can meet the performance requirement. For example, the buttenheads can develop the minim - guaranteed ultimate strength of the pre-stressire steel without failure er slip relative to the en:1 ancher.

Ecth the 90 and 170 vire end anchors met the deformation require-cent.

Tendons havire scre than 90 and less than 170 vires are possible without undue future testirg, if Only by using the 170 vire tenden end anchor with a lesse number of vires. Another performance requirement is that the end ancher be capable of =aintainirg in-tegrity for 500 cycles of loads correspondire te an average axial stress variation between 0 7 and 0 75 fs at a repetition rate of one cycle in 0.1 secc=d. ?nis require:ent, of course, sets ini==

acceptable limits en fatigue effects due to notchirg by the end ancher and tenden performance in response to ea-thquake leads. The g

U number of cycles was set by increasing to 500 frem the 100 predicted.

Tne stress variation was increased frc: a censervatively predicted 0.6 to 0.6L fi to the 0 7 to O ~5 fj. Further, the = ber of cycles caused by earthquake leaf. s predicted as enly 30 cf the total of 100 and by using all tM e streng ground =ctions which exceed one-half of the peak g-cun- atien for the earthquake. 7:e predicte:1 stress variaticn due to earthquake =ction alene is estimated as 10 percent cf the total of 0.0L f; stress variatien. The 0.0i f; predicted stress variation in turn results frc= cerbinations of '

earthquake, vind and accident leadings. Analyses =ade during the investigation inclu: led consideration of tenden excitation parallel and perpendicular to the tendon axis.

A ec parisen between perfc. ance requirements and tenden capabil-ities gives evi:dence that ths end anchers can satisfy the perfer:ance requirements.

The bearirs plate is include' as a pa-: cf the prestressing syster and is ene cf the greater interactions with the structure at its interface with the cen: rete. The average centact pressure between bearire plate and concrete is limited to these pe _issible by ACI 31o-o3 Tne maxi == contact pressure exceeds the average at Icea-tiens nearest where the end ancher centacts the bearing plate.

Tnis results free a -bending of the bearire plate. It is possible that the bending stress near the end ancher vill reach yield since concrete differential creep will increasc the bending of the plate j frc: its initially leaded condition, and the largest bending stress u

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O prior to shiptent, the vire is coated with a thin file of petrola-tu: centaining rust inhibitors, such as 'earbcrn Chemical Cc pany l No-Cx-Id h30 E cr 500. The intericr of the salvaniced, spiral-v spped seririgid corrugated sheathing is also 00ated with the sa:e este-

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a .* D=. a.-b . - N- a...d a.a'.. _- r. v ."w- Ox e

. . 1~ ^ C.F. *--e d .".e 7 . .' .' .' a.. . '."a. .e_""^..

and end anchors are surcunded by the No-Ox-Id which, in turn, is enca-sulated r by the sheathing and gasketed end ca-s e which are

.e e3_a_4 a e. *.e. " e '.a..do.. k ea. .' - ,la .a .

No-Cx-Id contains no solvents but does 00nnain certain proprie-tary che ical additives which inhibit cerrosien of steel. It is 4<1

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ing the sheaths and end caps it displaces air and water vapor be-

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the buttenheads.

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a=ounts of chlorides, sulfides er nitrates. However, to verify that none are present in the =aterial used, sa:ples are taken frc=

.'O r..-^e*+

a - , ^.' ea " N.' -a.,,

- ~ ** - v' " a. .'ea.e+. a. .ca= ,' .e r e . "ac-.

tory batch. The sa:ples are analyzed by an independent laboratcry as follevs: .

Water soluble chloride (Cl) is dete: cined-by AS24 Method D512-6.2T with a li=it of accuracy cf 0 5 pp .

Water soluble nitrates (N0 )2 are deter =ined by A324 Method D992-5,2 Water a.v, , , .v .. _e,._vith

. c. a-e.e a limit of ac':uracy ofv,0.01 =gwper liter.

( .e ) ....A .. ., < , .. . . . - - - , . a.c .n. 2 ..s.s-a . , n,)- .-<

~ . -

a _u 4 . n.

.r a .. ~~. a- ,.. , .

.o er .

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a

.. c '

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. .* *..* a. .=

• s a." .a.,ve d . r -^ a.. , +'

. - .na.

Dearocrn Chemical Cc=pany has stability data going back 18 years which indicate that the material does not deteriorate 'during the 40 years' life cf the plant. The chemical ec.positic: cf the = ate-4 a., , w...4 3.

. a,. b.~., . y,::,t p. ... .. ,e . ,,,e.. . ,a.a

. .- ., J. , 3,., 4

. 24.-a.. . .u. .. . . <. .

vould possess the ncr=al stability of liner hyi ocarbcns for the site temper =ture ranges. .

.h nea...

y....,. . . . . ><,4_4.e 1.4 _..

A research prog-a.: is now -*being -

conducted by 3echtel en a coating syster te

.va-la y -a. -

-- k e ^ .' .. a. .- e .' .c . =. . 1 k.e .~=. s -e n ..*.*...;*^. .=~.e.e . .v e . '..'..*.e. .ey .e. _=._ .' e.. sa-v.e . .a c.e -

. .e , a e"-

er .' a.-a. n . ". '. _" b a. ." ."o " e- * +~^ * *. .e ar"e.*-"4.~ -- N.a.q"e.e .'- s a.,

e w .amc.) v-4 .' +his re.. + mm A~.e.., 4 .e4 ec. < m.,.-.

The inside of the reactor building is lined with_a velded steel plate con-

~ fcr=ing to ASTM A-265, Grade A, firebcx quality consistent with the require-g

=ents for leak tightness. This steel has . a =ini=u: yield strength of 2L,000 psi and a =ini=u: elengation'in an 6 inch speci=en of 27 percent.

5-13 A=endment No. 5

.0 r 0,mq .m.4 << 11f.fgg .

The design, construction, inspection and testing cf the liner plate, which acts as a leak tight =e=brane and is net a pressure vessel, are not covered by any recognized code er specificatien. Hevever, the liner plate is sup-plied to the require =ents of ASTM A-235, " Standard Specification fer Lev and Inter =ediate Tensile Strength Carben Steel Flates cf Flange and Firebox Qualities for Pressure Vessels" and ASIM A-20, " Standard Specification for General Fequire=ent? for Delivery of Steel Plates for Pressure Vessels."

All ec=penents of the liner which.=ust resist the full design pressure, such as penetratiens, are selected to =eet the require =ents of Paragraph N-1211 of Section III, Nuclear Vessels, of the ASME Code. ASIM A-516 Grade c0 cr 70 =ade to ASTM A-300 is typical of a steel which =eets these require-

=ents and is used as a plate =aterial for penetratiens. This =aterial has excellent veldability_ characteristics and as =uch ductility as is obtainable in any ec==ercially available pressure vessel quality steel.

In acecrdance with ASME Code Case 13L7, allevable stresses fer A-516 Grades 60 and 70 are the sa=e as these per=itted fer A-201 Grade 3 and A-212 Grade 3, respectively.

The A-285 =aterial was chosen on the basis that it has sufficient strength as well as ductility to resist the expected stresses frc= design criteria leading and at the sa:e ti=e preserve the required leak tightness of the reacter building.

s The liner plate vitb' the exception of penetration einforce=ents is designed to function enly as a leak tight =e=trs=e. It is not designed to resist the

~

tensien stresses frc= internal applied pressure which =ay result fr~n any credible accident condition. The structural integrity of the reacter tuild-ing is =aintained by the prestressed, post-tensioned concrete. For these reasons and because the liner plate is pri=arily in cc=pression, the deter-

=ination cf nil-ductility-transition te=peratures as specified in Paragraph N-330 cf Section III, Nuclear Vessels, cf the ASMI Ccde does net apply to the liner plate. On the other hand, all =aterial for reacter building parts such as penetrations, which =ust resist applied internal pressure stresses, is i= pact. tested in accordance with the require =ents cf Faragraph N-1211 cf Section III, Nuclear Vessels, cf the ASME Code.

A-265 steel is readily veldable by all of the ec==ercially available electric are and gas velding processes.

A funda= ental require =ent for fabrication and erection cf the liner plate is that all velding procedures. and velding cperators te qualified by tests as specified in Secticn IX cf the ASME-Cede. TP's Code requires testing cf velded transverse rect a=d face' tend sa=ples .n ' crier te verify adequate veld neta11 ductility.

~

Specifically,Section IX cf the Code requires that transverse rect and face bend sa=ples be capable of being bent -ecid 160 degress te an inside radius equal. tc twice the thickness of the test sa=ple. Satisfactcry ec=pletien-cf-these bend tests is accepted as adequate evidence of required veld.=etal and plate .=aterial ec=patibility.

5-lh Q ^rD'/g A=end=en+

i i

(/T

\'- Mill test results are obtaired for the liner plate material. The plate is visually checked for possible laminations oad pitting.

The surfaces of the liner plate not to be in contact with concrete are protected by an initial surfact cleaning and a prime coat of paint applied at the fabrication plant to protect it until installation. A painting system compatible with accident conditions is applied to the exposed surface after the plate is installed.

5.1.1.4 Reactor Building Design Criteria Safety of the structure under extraordinary circumstances and performance of the reactor building at various leading stages are the main considerations in establishing the structural design criteria.

The two basic criteria are:

a. The integrity of the reactor building liner is guaranteed under all credible loading conditions and,

b. The structure has a low-strain elastic response such that its behavior is predictable under all design loadings.

The strength of the reactor building at working stress and overall yielding j

is compared to various loading combinations to insure safety. The design of the reactor'builcing is anslyzed with respect to strength, the nature and the amount of cracking, the magnitude of defor=ation, and the extent of cor-rosion to insure proper perfor=ance. The structure is designed to meet the

. following conditions:

a. Prior to prestressing

b. At transfer of prestress

c. Under sustained prestress

d. At design loads I'

.e. At factored loads Deviations in allowable stresses for the design leading conditions in the

, working stress cethod are per=itted if the factored load capacity criteria.

l are fully. satisfied. All design is in accordance with the ACI Code 318-63 unless otherwise stated herein.

No special design bases are required for the design and checking of the base slab. It acts primarily in bending rather than cembrane stress. Design of the 26 base slab is based on the ACI Code 318-63, except that the maxi =u= reinforcement spacing of the base slab is increased to 30" due to the massive depth.

The loads and stresses in the base slab, cylinder, and dome are determined as

'h described below.

. [V O O M .9 Amendmant No. 26 5-15 4/74 i

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,/ X

_s P Pressure load varies with time frc: design pressure to ne pressure.)

2, . T.ha.- c .' 7~a d s

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the vessel to eliminate membrane tensile stress (tensile stress across the entire vall thickness) under design loads. Flexural tensile cracking is per=itted but is controlled by bonded reinforcing steel. The control of cracking under operating coniitions denoted by Equation (a) above is in accer-ac,c

--- . v4..x

.- .3.

w . --ov.4

e. sic - . w,n+x .-' U s.S- 6.1, 4ea-4-- ---- -p'

' .OaR.

Under the design loads the same performance limits stated in 51.1.k.3 apply with the following exceptiens:

a. If the net =edbrane compression is below 100 psi, it is neglected and a cracked section is assuned in the ensputation of flexural bonded reinforcing steel. Flexural tensile stresses in bended reinforcing cf 0 5 fy are allcaed.

b. k#nen the re.xiru: flexural stress does not exceed 6Y f,'v and the extent of the tension coua is no =cre than one-third the depth of the section, bonded reinforcing steel is provided to carry the entire tensien in the tension block. Otherwise, the bonded i ,}'

(

reinforcing steel is designed assuring a cracked section. When the bending =0:ent tensicn is added to the ther=al tension, the allovable tensile stress in the bonded reinforcing steel is 05fv =inus the stress in the reinforcing due to the ther=al gradient as deter =ined in acecriance with the =ethod cf ACI-505

c. The prcble of shear and diagenal tension in a prestressed con-crete structure is censidered in two parts: =ebbrane principal tension and flexural principal tension. Since sufficient pre-stressing is used to eliminate membrane tensile stress, membrane principal tension is not critical at design loads. Me=brane principal tension due to combined =erbrane tension and =erbrane shear is considered under 5 1.1.h.6.

Flexural principal tension is the tensicn associated with bending in planes d d ~"..' a .- .*c .ha. s" .#a a. a. ..

e .-ny a. --. .

^*-- 'ka. .cha..'.' and . .=haa-

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.e .. - *.*a-shell (radial shear stress). Ihe present ACI 315-63 previsions cf Chapter c;e r-en,.-.

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as discussed under 5 1.1.k.6, using a load facter cf 15 fcr shear leais.

Crack control in the concrete is accc plished by adhering to the ACI-ASCE Code Oct:ittee ctandards fer the use cf reinfcrcing steel. These criteria are based upon a recc=nendation of the Prestressed Concrete Institute, and are as follows:

5 5-18

{g 0.25 percent reinforcing shall be provided at the tension face for s=all =e=bers.

0.20 percent for =ediu: size =e=bers.

0.15 percent for large =e=bers.

_A =in1=u= 0.25 percent =ild steel reinforcing vill be pro-vided in two perpendicular directions on the exterior faces

, of the vall and do=e for proper crack control.

Since,'in general, there is no tensile stress due to te=perature on the inside faces of the reactor ou11 dings, =ild steel reinforcing is not neces-sary at the inside faces.

5 1.1.4.6 Factored Loads The reactor building shell is checked for the factored loads and load co=bina-tions given below. t The load factors are the ratio by which loads are cultiplied for design purposes. The load factor approach is used in this design as a =eans of

=aking a rational evaluation of the isolated factors which =ust be considered

in assuring an adequate safety =argin for the structure. This approach permits the designer to place-the greatest conservatis on those loads =ost

~

subject to variation and which =ost directly control the overall safety of g the structure. It also places =ini=ur e=phasis on the fixed gravity loads

' and maxi =u= e=phasis on accident and earthquake or vind 1cais.

The final ~ design of the reactor building satisfies the folleving load ec=binations and factors:

a. C 1/p (1.05 0 A 1 5 P + 1.0 TA + 1.0 F)

~b. C 1/p (1.05 D + 1.25 P + 1.0 TA+ E * *E + *

c. C ~ 1/% (1.05 D + 1.25 H + 1.0 R + 1.0 F + 1.25 E + 1.0 To )

d. C 1/p (1.05 D + 1.0 F + 1.25 H + 1.0 W + 1.0 Ig)

e. C 1/p(1.0D+1~.0P+1.0T + 1.0 E + 1.0 E' + 1.0 F)

f. C 1/p . (1.0 D + 1.0 E + 1.0 R + 1.0 E' + 1.0 F + 1.0 T )

g. C 1/9 (1.0 D + 1.0 M + 1.0 F)

Where:

C = Required capacity of the structure to resist fcetored loads.

@ = Capacity reduction factor (defined in 5 1.1.4 7).

n .

s./ \

c

( }

._Q/

19

4 i

, s

, D = Dead loads of structures and equipment plus any other pez ::anent load contributing stress, such as hydrostatic or sci]. In addition, a l portion of the live load , added when it includes items such as piping,

, cable, and trays suspendea frc= floors. An allowance is made for future additional per=anent loads.

f P = Design accident pressure load.

! F = Effective prestress loads.

.R = Force or pressure on structure due to rupture of any one pipe.

i H = Force on structure due to thermal expansion of pipes due to design conditions .,

} T = Thermal loads due to the temperature gradient through wall during opera-ating conditions.

TA = Thez=al loads due to the te=perature gradiet.t through the wall and expan-4 sion of the liner. It is based en a temperature corresponding to the

- design accident pressure.

t E ." Design Earthquake" load.

I' E' = " Maximum Earthquake" load.

W Tornado load.

i l M = Maximu= probable flood load. .

4 Equation (a) assures that the reactor bu11di 6 has the capacity to withstand

' pressure loadings at least 50 percent greater than those calculated for the postulated less-of-coolant accident alone.

Equation (b) assures that the reacter building has the capacity to withstand i loadings at least 25 percent Ereater than those calculated for the postulated loss-of-coolant accident with .a coincident design earthquake.

i Equatica (c) assures that the reacter building has the capacity to withstand

] . earthquake loadings 25 percent greater than those calculated for the design earthquake coincident with rupture of any attached piping.

. Equation.(d) assuzes that the reacter building has the capacity to withstand a tornado' loading.

Equations (e) and (f) assure that the reactor building has the capacity to

t. withstand either the postulated loss-of-coolant accident or the rupture of any attached pipin6 coincident with=the maximum hypothetical earthquake.

Equation (6) assures that the' reactor building has the capacity to withstand maximu= probable flood.-

[)v) The stress in prestressing steel and bonded reinfercing steel is . limited to f,wherefgistheguaranteedminimuryieldstressgivenintheappropriate y

i

5-20 00084 j '

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+ ,.< a- c, . , s m. a- er .v .h.

--. m.s a2 < oc._ e-. w e-,

+

.enoa ..

Principal seibrane tension in concrete is calculated due to coibined terbrane

• e

. .e.t^n. an d -a_-t.a-a.

_ - .e k a. a *, a x. .-

' " i d -e '_'

. -- . a. x." -c - ..' *a. =-io- due *^ be-*_* _. e- _ _-a__ e e

er thermal gradients. When the value cf principal cembrane tension stress exceeds s+.e e., .e eis s 3'/.f,', the,acorbination _, c,-_, a. e 4 .a_, .

of banded reinforcing steel and prestressing

. . .n.

. .o t .4,.,.<,_, y _---a-b ra"- a. .

• a.ns.-- 4 ^* . v'

• n' s-" *.

ex-aad.' 3 -

• "a.

.. ab ^va. -a

- *.* .^ na.d .e .. + -a. .c .e _'.4-.'*.u**.*** .- ..

^# .#

}. .

.+..

n--- .h. . . va.,,,.. . v,, ..

y.<.n , .< gu, ., -_ m .w. -- e . o.. e- e _< .- a.

- - mo s-m -. .. x._e. 2. .: ,icn.e. -

an m. -

• ha.

. - _4-

e. - 4pa' o--.- .-e+a. *a- .--o .*w--

d ua. - c w _-k .--- 4 a d m- a ~-k--.-e - a.

• a..s.io- . , _C=_k.

a- a shear, and axural tension due to bending morents or thermal gradients o.x , e,. a *

• s 6Y ".,', bcnd

. - ed- . -a.4 -... . #^*c4n- e.e*ea..' s p.wv4 dad -- -

.--- *ka. .w.

.~ #^'_'o.d s- c-- -- na..-.

O. e. , , r _e m xu, .,,.,

e < n- . E a n a. ..e< . . .e . .< e s,.e,_, _s y.. o

u. m. e- .- < a s _aea

<- annm.a

.- .. . a uen. e . a.. . .-  %+% .u a. -a..%

. .u mas c. Ari.50).

o

. 7.% a. -.-t e .< _-n.- _ a ,a

. m~.r s *.e a ' p.o.4ded . - - 4s 0.2p 'ye--a. .. - +. 4- eack dd-ac+.'^

7 ~s . . .

t, ,

xj a ,3

s. M n-.< -

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.c

. .a- -n._-a__. v- .u. w

- . n v .o c . nu.e n

.r sn.-+.<w-- -

a

.w

. ... ,7 ,,- - a 4.. .g .h . . 3.e_, - .a . s...es

. . . = e.a.m.a

. . abe,s,. . ..a..w .a .w. .. < .,.,

.. v. .4.g -

, x .a.,. < n - .

r.---.

.n- m a. .. .n.

. ,u e -

a_<_e . ,_-n.n.. e e=i,__a

- e o a a.a

-a n -.-+.,...__e.

> .na +,-s_<-,

e.- -

• --k a. a ll^". ak..i a. *. a. s _4 .' e s .. a. .e .=' .- - *u"a. . -a. -. .# .# . .

• 3 .e . a. a. _' .d .- . y - - "-

- -d

- .e

• he

. .e+.*ess .4-- .

4

~. #.^ .= .a .-3' n - A - o.,. *

. o +.k

-- a .. +ho. a ., .a +. a.e u *.e _.dnc.a.

e--a-_<o. 4h in accordance with the methods of ACI-505 ck.ea.

-. s+.-a.

. e e _' d_--1 +. .e a --' .ek a.a- . . -a..' n'^v . . . .a d e .'r.- .adda' . .ekaa a-a.- _d-_

^' e 0

m" a - ~ ^.-da m a a. . . . - u- = r"*. a."

.. . .. ACT . .U. c: . .4 *.-b . " e .'o _' _' v v' _s ex ^ e.r-_

+<c

.. - e

.r.. _- ..e ,

ob' ' . . ' .5. c. ^oda. . - .e.-k.a _' '_ b a. .e*_- ,'c^ =.4 kv v.

U 9,. ,,

= :.* k.fa- .e, t

+ yr a,. (.,b -, ) e i

4 ( _' )

yu .n ., n_ ,

f 0.026 ,

+ . a, av. ,

z. - 1.tv - -,,

-r

(~.;

(1)This fcr=ula is based on the recent tests and work done by Dr. A. H. Mattock

_j; .P

^. . + h,. U A. srn. c... 4 .s,. . A .P TnTa.e .% .<

v e-..e-m .

5-21 0 05. % Arendment ho. 5 11/3/69

4 1

i 1

1 y but not less than 0.6 for p' '> 0.003 For p ' < 0.003, the value of K shall be cere.

.. e .

M Or

=IY 6 Yf: ' + fpe +f n

+ f, f P~, = Cc=pressive stress in Ocncrete due :: prestress applied

=cr=al to the cross section after all losses (including the stress due to any secondary =c=ent) at the extre=e fibre of the section at which tension stresses are caused by live loads.

i. f,,. = Stress due to axial applied loads (f. shall b'e nen.tive for . -

tensien stress and positive for c0=pressic: stress).

4 v

, f, = Stress due to initial loads, at the extre=e fibre of a

^

section at which tension stresses are caused by applied

loads (including the stress due to any se

Ondary =c=ent).

f, shall be negative for tension stress and positive for c0=pression stress.

4 n , 505

<Mr f

, s-i T*e '

1.

N

p. ,.A' s bd V . _ Shear at the section under censideration due :: the applied

. ,.o- ae.

M' .. Mc=ent at a distance d/2 fr0= the .section under c:nsidera-tion,.=easured in the direction of decreasing =0=ent, due to applied loads.

V, . Shear due to initial leads (positive when initial shear is

^

in the sa=e direction as the shear due to applied leads).

IcVe

• limit'placed by ACI 313-63 en c' v , as 17 YdQ' '

is not

, ' applied.

7:=:ula 26-13 cf the Code shall be replaced by

.. r r-

..CV = 37- b,2u yf..Y1+'pc+*n C:

a

t. )

- c ,.

3v -

k The ter= fg is as' defined above. All other notations are in-accordance"vith Chapter 26, ACI 315-63 R '

~/

D') .' (2)This for=ula is based on. the ec==entary for Proposal Redraft of Section 2610,

' ACI-318 by Dr. A. H. ' Mattock, dated Dece=ber 1962.

22 00086

. ,. -= - ~ , - . . . - , - . - . . . - - . , . ,- -.a-., - - - . ,

_. _ . -- . - . _ ~ - - _ _ . -. _ -. _

i F

1

( When the above-mentioned equations show the allovable shear in concrete is cero, radial horicental shear ties are previded to

, resist all the calculated shear.

5 1.1.4.7 Yield Capacity Reduction Facters The yield capacity of all load carrying structural elements is reduced by j- a yield capacity reduction factor (%) es given below. The justification for these numerical values is given in Appendix 5-E. This factor provides for "the possibility ;nat small adverse variations in =aterial strengths, 4 verkmanship, dimensions, control, and degree of supervision, while individ-ually within required tolerances and the limits of Sood practice, occasionally

_~

ray co=bine to result in undercapacity." (Refer to footnote en Page 66 of t

4.CI Code 318-63.)

! Yield Capacity Reduction Factors:

i p = 0 90 for concrete in flexure.

p: 0.85 for tension, shear, bond, and anchorage in concrete.

p = 0.75 fer spirally reinforced concrete co=pression members.

p = 0 70 for tied compression =e=bers.

i

()

\- /

'O = 0 90 for fabricated structural steel, p: 0 90 for reinforcing steel in direct tension.

Q = 0 90 for velded er mechanical splices of reinforcing steel.

Q : 0.85 fer lap splices cf reinforcing steel.

.% 0 95 for prestressed tendons in direct tension.

5 1.1.k.8' Prestress losses 4

. In accordance with ACI Code 316-63, the design provides for prestress losses caused by the following effects:

i

a. Seating of anchorage.

(

l - b. Elastic shortening of concrete,

c. Creep cf concrete.

d. Shrinkage of concrete.

e. Relaxation of prestressing steel stress.

P

f. Frictional loss due to intended er unintended curvnt".re.in the s

tendons.

i

%,l ' 'All of' the above losses can be predicted with a. reasonable de6ree of accuracy.

>.-23 .

00087

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s

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24r.c..a .,

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...-e w.. . 4 ., 2 4 _ e.

gener...14 a.4me.e. ...- .a r

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.- g o. n .' .c .. a. +.-b a. g v. .u. s.

. ~ --- nea L.une a sm- .4 , a V.1 a. .m 4+ *

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. - -.. . r....

=issiles. (See 5 1.1.L.10)

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. .. = a. .g m y.a. ...-

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.4mo. mnA

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. Aa. . s.e. o.vwe n

- a..

e a-A--

e ne.a.m, N .4*p s +4

.. -ems +O .4c.4

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p .e.e..e ve.e.ee.,,,

.. ...e.e ..

r... a r.4 . r .4- w

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r. -

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• k a. e.' .* *_ea

..* - * ^ - .s,ta 4 .,~ .."*..ba.

^

• • a. . .da.'. ( .r.^ -

I tild steel -at 4 -

failure, this elongatiCn varies frc: 15 percent to 30 percent. )

e. go. .e.. .c.4,.e .o-..,*

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3 ' o m b.' a. .-a.=. *

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- c ' +." a. n c.vr .Ec.' .' a.*. a -A P. a. . .. e". e Va. .e s a. .' code, c -+4^-e .. as

-  ??.?, ." ^. .' a. a "v e .e .= .e ' .e , n.*..

• * . .'e L. ema ar . a .d.'.'. ~.a .' .' y, *. h e

.c ,~. 4 -

• e , , . . , 4s~ .e a. a.. *. 4 ".- e have baa- ad..r-* e 4 -c .e e"a.d-a. .e -- 4 e .e

• aS..' .d .e b .d - - a' '.n. ;a k. .' a.

. -s _

...a e * *e . m - .~ ... .4 + e .

.. c c.w r -. . . .e r.L,.

a .

(-)

- e.va. --

e-

.e....a. e .

, e (t

e- a.a-e. r-s..'.~.,c

-e r-tn. e . ....e.- = . . . r. .,.. ..., n. .o. ....,

nw

-. . u. r, . .' .. ,;

Figu es "-kik, N kl5 (A)

Paragraph N-k12 (n)

Paragraph N h15 1 m

\

t 1

-%s 5-2L @%S A=end=ent No. 5 11/3/69

. - - . _ . -. - ._ . - _ --~ - - _ . _ _ _ _ - . - . . __ _ -__ _ _ _ _

1 i '

t 4

(^ '

I=ple=entation of the AEG design criteria requires that the liner raterial be prevented fro: experiencing significant distertic due to the ther=al load and that the stresses be c ~ *d - ad ~~ a faticae standpcint. (para-graph N-k12 (=) (2))

t The fellev1cg fatigae leads are considered in the design cf the reacter i building liner:

i a. Tner=al cycling due to annual cutdo:r te=perature variatices.

Daily te=perature variations dc =0t penetrate a sig0ificant dis-tance into the 00ncrete shell te appreciably change the average te=perature of the shell relative to the reactor building liter.

The nu=ber Of cycles for this leading vill be hO cycles fcr the plant life of LO years.

b. Ther=al cycling due to reacter building interior a=perature varying during the start-up and shutdevn of the reu 'tcr syste=.

The nu=ber cf eyeles fer this loading is assu=ed to b; 500.

c. Tner=al cycling due to the less-cf-coolant accident (IOCA) is assu=ed to be ene cycle. Themal lead cyc_es in the piping systers are s0=ewhat isolated ^ ^ - +he reactor building liner penetrations by the concentric sleeves between the pipe and the liner. The attach =ent sleeve is designed in at:Ordance with ASG Sectica III fatigae consideraticus. All penetraticus are reviewed for a p ecaservative nu=ber of cycles to be expected during the plant life.

The ther=al stresses in the liner plate fall into the categ0 ries considered in Article .k, Sectic III, Uuelear Vessels of the ASE Boiler and Pressure Vessel Code. The allevable stresses in Figure N-415 (A) are for alternating

, stress intensity for carben steels and te=peratures n0 exceeding 700 F.

In additice, the ASE Code further requires that significant distertio: Of the =aterial be prevented. t In accordance.vith ASE Code Mgraph hl2 (=) (2), the liner plate is re-strained against significant distertion by continuous angle anchors and never exceeds the te=pera*"~ '* *tation Of 700 F and also satisfies the criteria for li=iting strains on the basis of fatigue consideratic .

paragraph k12 (5) Figure N-415 (A) of the Asc Code has been developed as a result of research, industry experience, and the proven perfer=ance Of 00de vessels. Because Of the conservative facters it contains concerning both stress intensity and stress cycles, and because it is.a part cf a reccEnited design code, Figure N-h15 (A) together with its appropriate li=ita:10:s-has been used as a basis for establishing allevable reacter building liner strains. Since the graph in Figare N-415 (A) d es act extend belev 10 cycles,10 cycles are used fer an IDCA instead of cne cycle.

Establishing an allevable strain based on the one significant ther=al cycle ci the accident condition would per=it an allovable strain (fr0= Figare p .N L15 .(A)) cf approxi=ately 2 percent. The strain in the liner plate at O

5-25 00M9

.. . ~ _ . _ . . _ _ _ _ - _ . - _- __ __

i J

4 l its proportiona' '*~** *s approxi=ately 0.1 percent. The reacter building l liner is allowed to go beyond propertions' ' d-4

• strains during the accident condition. Maximu= all vable tensile er c0=pressive strain has been con-servatively set at 0 5 percent. The =axi=u= predicted =e=brane strain in a

the reacter building liner during accident Conditions has been found to be I 0.25 percent. The =axi=u= c:= tined =e=brane and flexural strain is pre-j iicted to be O.k5 percent. At the design accident pressure condition, there j '

is no tensile stress anywhere in the reactor building liner =e=brane. This is true both at the ti=e of initial pressure release and later, under any i accident pressure and te=perature 00niitics. The purpose cf specifying an NDT te=perature require =ent is to provide protection against a brittle frneture er cleavage = ode of failure. E0vever, this type of failure is j pre luded by the absence cf tensile stresses.

i

. No allovuble ec=pressive strain value has been set for the test condition

!, because the value is less than that experienced under the accident condi-tions. The =axi=ws predicted ec=pressive strain is approxi=ately 0.07 per-cent.

The maxi =u= allovable tensile strain is 0.2 percent under test conditions.

The predicted value vill be very nearly zero.

The stability cf the liner plate is insured by the stiffening and anchoring of the plate to the prestressed concrete structure as is indicated by the l typical details shown in Figure 5-1.

C i ( The maxi =c Oc=pressive strains are caused by accident pressure, ther=al leading, prestress, shrinkage and creep. The =ax1=u= strains do not exceed 0.025 inch / inch and the liner plate always re=ains in a stable ceniition.

The conservative design approach of the stiffening syste= used in the liner

. plate to prevent significant distortions at accident condit10ns, and strin-

, gent velding and weld inspection require =ents insure that the leak tightness l cf the liner plate at accident conditions vill not change frc= that at the

! test conditions.

The =0st critical condition for a liner plate exists when the liner plate is in the condition illustrated by Figure 5-6. In this condition, Panel 1

! and Panel 3 have outvard initial curvature and Panel 2 has inward initial curvature. When a lead is applied to the liner plate, Panels 1 and 3 bear against the concrete and Panel 2 defor=s inward. If the load is pri=arily fr0= conprete shrinkage, creep, prestress and ther=al effect3, t stress (j) in Panels 1 and 3 tends to relax to a value of ( "-b,hei =e=brane

) in

!. panel 2. The anchers between the panels with inward and outvald curvature nust restrain a force cf AU fer stati: equilibriu=. Due to inward deforma-tioL, flexural stress also exists in Panel 2 and the anchors are subjected l to the =o=ent (M). (See Figure 5-6.)

Due to the fact that all of the significant c =pressive liner plate loads

-are self-li=iting, ie, as deformations occur, the loads tend to reduce; the

, liner =ay defor= inward but can never get into an unstable condition provided 1

it is sufficiently restrained.

00090 5-26

p g\~-) The only significance of the rathematically calculated elastic buckling stress (e c.;.) is as follevs: If a panel is perfectly straight and (c ,,_)

ic lover inan the yield stress (c y), then when ( CR) s reac e , hie" panel vill tene to defert inward i: an internal pressure is not present.

When a panel has initial inward curvt.ture the rate cf change of inward deflection with respect to renbrane stress vill increase after (c g-) is exceeded, but the panel v -a-ain stable. If a perfectly straig59 panel has (c cc.). higher than ( c /), then the panel vill defor inward when (; v),

.s reacnea.

• The anchor detail has the capability of resisting the full force (aN) due to a theoretier.11y fixed anchor, but in addition, it has sufficient ductility' to accept the .6087 inch displacement without failure. The above displace-cent results from a unifer: cembrane strain of .0325 inch / inch distributed over a 14 inch anchor spacing. The 1k inch ancher spacing is preli=inary; this spacing may be increased as long as the structural integrity and leak tightness of the liner vill be maintained. Variou.e patterns of velds attach-ing the angle anchors to the liner plate have been tested for ductility and strength when subjected to a transverse shear load such as AN and are shown in Figure 5-7

1. Panels 1 and 3 are thicker than Panel 2; the feree across the anchor vill increase; the thickness variation for a plate can be obtained fcr " ASTM Part L, January 1967, Specification A-285 and Specifica-tion A-20-66" which states that the tolerance is +16 percent and L percent for 1/L inch thick plate.

p

( 2. An increase in yield strength cf the liner plate =aterial vill in-crease the force across the anchcr. The information vill be obtained rrc= the " Mill Test Reports." For preliminary design values, a 50 percent increase vill give adequate protection; this vill result in the yielc stress increasing frc= 2L ksi to 36 ksi.

3 A variation of the todulus cf elasticity and Poisson's ratio does not appreciably affect the design er the margins available due to the ductility of the anchorage syste=. The design vill give adequate protection with respect to this variation.

The above infer =ation is preliminary and is based en the use of A-255 plate.

By keeping the anchor spacing small, the a=ount of force LN and the invard displacement =ay be controlled. A =ax1="- ***+'al inward displacement be-tween the anchcrs of 1/8 inch is also specified and contrelled in the field.

This value is used in the design since it is important in co --"4"g the amount of inward displacement and the a: cunt c f relaxatien in Panel 2.

i A -

\

00iM1

5-27 .amendent No.5 11/3/69

m i

y ,4 O #. f.e a.*. s a *. .' .* - e. r* .i a *. a. .e e .n .e a .-a. c .. ^ - *. .*^. .' .' a. d .d - a-~..~.~'a.^.=.

. . ".d*x..- nSyr. ca.a*'--

t r.y n. o d o. , v wa .4-w a.' .' ov.e .' ./

~. 'e a .. '*- s -- a. - - *. #^- .'

.k r .' a *. a. . .. a.

- ... ~

.' .. ~~

flexural strains due to the :: ent (M) are added t: calculate the total strain in the liner plate.

. ~w . ..' n a.r.- ' at e ' s a-- ~..k.. .-a. ' --

c-

.e h^w

. ' n .v4 s-"a. e ,R

. v.' *.* - -a ~k ..-a c_a. d. . . .* .^ .* " .

.w

. ~ e 1, -s .* *."- ' i na ' -- a ' '--

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b-g the designer for the appropriate plant design basis.

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and anchcrage design together with the effects Of shrinkage and creep cf acnarete and vacuu: loads. The penetration assemblies are designed to aa^^.s_- ^iata. s.'.'^.'*."a. ab va. .' -sad -e v.' a. .-a. a-rr .' .d a.ak..' e a---d a .' .e .- *k - e e . .#a.'

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. c + e ex e.e.e.' ve strains at the disecntinuities. The forces in the liner plate at the dis-0:ntinuities are evaluated by use of the finite elenent =puter program and the anchors are designed to resist these forces.

At all penetrations, the reactor building liner is thickened to reduce stress con:entrations in acecriance with the ASL2 Boiler and Presst7e Vessel L de 1965, secti:n III, nu: lear vessels. The thickened pertien of the liner plate is then anchered to the concrete by use of anchor studs : pletely around the penetrations. For details of the penetrations see Figure 5-2.

3 The sleeves, pipe cap, and all velds associated with the penetrations are

%d

) designed to resist all loads previously =entioned and also the prestress ,

ferees and internal design pressure.

5.1.1.k.10 Missile protection criteria

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k'm-) 5 1.1 5 1 Critical Design Areas The :ritical design areas of prestressed concrete reactor buildings are:

a. The restraints at the top and bettc= of the cylinder.

b. The restraints at the edge of the spherical sector dO=e.

c. The stresses around the large penetrations.

d. The behavier of the base slab relative to an elastic foundation.

e. The stresses due to transient te=perature gradients in the liner plate and concrete.

f. Stresses within the ring girder,

g. Penetrations and points Of concentrated load.

3 1.1 5.2 Analytical Techniques The reactor building analysis is perfomed by the finite ele =ent =ethod developed by E. L. Wilson, under sponsorship of National Science Foundation Research Grant G189o6. This progra: has been further developed to apply tc axisy==etric structures. Such a method of analysis is nor= ally used enly for thick walled structures where conventional shell analysis yields

[ml v"/ inaccurate results. Good correlation has been de=cnstrated between the finite ele =ent analysis =ethod and the test results for thich vall =odel vessels.

The design analysis for axisy==etric icadings such as dead load, live load, te=perature, and pressure is done fer Ite=s (a) - (f) using the finite element ec=puter progra= because all of the conditions are axisy==etric. Other loadings such as seis=ic, vind and concentrated loads which are nonaxisy=-

=etric are handled by techniques described in 5 1.15 5 and 5 1.15.6.

The finite ele =ent technique is a general method of structural analysis in which the continuous structure is replaced by a syste= of ele =ents (=e=bers) connected at a finite number of nodal points (j0ints). Conventional analyses of fra=es and trusses can be considered to be exa=ples of the finite element nethod.

In the application of the method to an axisy:::etric solid (eg, a concrete reacter building), the continuous structure is replaced by a syste= cf rings Of triangular cross section which are inta--= ed alcng cir~~ca~ -w joints. Based on energy principles, work equilibriu: equations are formed in which the radial and axial displace =ents at the ci"~ ~#a~ ntial joints are unkncvns of the syster. A sclutien cf this set of equations is inherent in the sclution of the finite ele =ent syste=.

The finite ele =ent grid of the structure base slab is extended down into the m x foundation r.:aterial to take into consideration the elastic natu e of the

(

s ,

j foundation material and its effect upon the behavior of the base slab.

v 5-30 000M .c-e

/'

N t I The use of a finite ele =ent analysis permits an accurate dete=ination of G the stress pattern at any location en the structure. Tne analysis method has been de=0nstrated on the following types of structures:

a. Arch da=s (including a pertion of the foundation),

b. Thick valled prestressed concrete vessels.

c. Spacecraft heat shields.

d. Rocket neccles.

The ec=puter progra: used in the analysis handles the fellowing inputs:

a. Twelve different =sterials,

b. nonlinear stress-strain curves for each =aterial.

c. Any shape transient te=perature curves,

d. Any shape axisy=setric loading.

The progra= outputs are:

a. The direct stress and shear stress for each ele =ent.

I,,'n b. Tne principal stresses and their directions for each element.

\v/

c. The deflections for each nodal point.

An auxiliary ec=puter progra: plots isostress curves based on the above analysi progra: cutputs.

Additional infor=aticn regarding this technique, the co=puter progra=

e= ployed, and a c0=parison of the results with other analytical =ethods are contained in Appendix 5-F.

51.153 Ther=al Icads The thermal loads are a result of the temperature differential within the structure. The design temperature gradients for this structure are shown en Figure 5 h. The finite element analysis was prepared so that when ter-perat res are given at every nodal peint, ctresses are calculated at the center of each ele =ent. Tnis way the liner plate is nandled as an integral part of the ctructure, naving differen: =aterial properties, and not as a techanis which would act as an outside source to produce leading en the concrete pertion of the structu e.

The liner plate is designed to have plastic defc=ation as a result of pre-stressing and high ther=al stresses. Tne finite ele =ent =ethod includes this analysis too, by successive approx 1=ations, changing the modulus of elasticity of those ele =ents which are subject to stresses higher than the

[_';/

s propertional li=it.

i/

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5-31

The output of the co=puter analysis shows the effect of the thermal loads on liner plate and concrete. Under this condition the liner plate and the in-side of the cenerete are subject to ec::pressive stress and the outside of the concrete section is subject to tension. These tension stresses balance the co=pressive stresses so that, except close to any discontinuity, there is no resultant =e=brane force. That is, all the co=pressive forces in the liner plate are carried by the prestressed concrete and reinforce =ent near the outside surface of the structure.

i The cc=pressive stresses in the liner plate exceed the propertional limit in the case of the design basis accident. An increased temperature would keep the liner plate in plastic condition, but only a negligible additional stress could develop, and tiemal stresses vould stay unchanged.

1 The small increase in reactor building te=perature to that temperature associated with the factored pressure is not significant in changins the te=perature gradient through the concrete. Therefore, the temperature gradient as shown in Figure 5-4 is used for both the design accident and factored accident.

5 1.1 5.4 Tendon Failure Analysis l There are approxi=ately 110 vertical tendons, 90 dome tend.ons, and 165 hoop tendons. The hoop tendons are placed in 2ko degree sections around the cylinder using three buttresses as anchorages.

m All prestressed tendons are subject to the =ost critical stress during ini-tial tensioning. There vill be a loss of prestress on the order of 15 per-cent due to elastic and plastic losses, which vill reduce the stress to the design level. Even at the factored yield loads, the stress in the tendons is not as high as during initial tensioning. Each of the tendons has been pretested at the ti=e of initial jacking and the stress in the tendons under accident loading is approximately 80 percent of the jacking stress. This means that the possibility of tenden failure under design accident loading is quite remote.

Although there is ample reserve capacity in the tendon and structure, the i

co= plex nature of the structural behavior makes it difficult to predict the effect of a hypothetical series of tendon failures until the final design .

is co=plete.

If two or three of the tendons fail during accident conditions, and if they i

are side by side or close together, this will not affect the integrity of the structure er the liner because the thick concrete valls vill be suffi-cient to trans=it the force from the adjoining tendons without resulting in any serious local stresses.

5 1.1 5.5 Stresses near Equip =ent Hatches 51.1551 Analytical Solution: Analytical solutions for the deter =ina-tion of state of stress in the vicinity of the equipment hatch are obtained from reference to the following article: (a) Eringen, A. C.; Naghndi, A. K.

p and Thiel, C. C., State of Stress in Circular cylindrical Shell With a Circu-lar Hole, Welding Research Council Bulletin No. 102, January 19o5 5-32 GPS Amendment No. Ik 7/31/70 i __ . _ - - - -. --

A

't

-\ The analysis of the reacter building as a whole is first carried Out without considering the Openings in it. This analysis is done by using the finite element program.

The reacter building with the opening in it is then analyzed in the felleving steps in a verking stress (elastic) analysis:

1. For=ulation cf differential equations for the sha ** ~~= plex variable for= with the center of the hole as the crigin. (See Reference (a) above. )

2. Solution of the differential equations. (See Reference (a) above.)

3 Evaluation of para =eters in the solution. (See aeference (a) above . )

k. For=ulation of the boundary conditions based on the stresses obtained frs= the vessel analysis above without the hole.

5 Calculation of =e=brane forces, =c=ents, and shears around and at the edge of the opening.

6. The vall thickness around the cpening is then increased at the outside face.

At the equip =ent hatch, the reacter building vall is thickened for the folleving reasons:

S s s, a. To reduce the larger than acceptable predicted stre:ses around the opening.

b. To acec==odate tendon placement,

c. To accO:=odate bonded steel reinforcing place =ent.

d. Tc ec=pensate for the reduction in the overall shell stiffness due to the opening.

T. Evaluation of sc=e of the effects of prestressing that are not handled in Reference (a) above.

S. Finally, the design is checked to insure that the scrength of the rein-force =ent provided replaces the strength re=0ved by the cpening. This check is to maintain a good degree of c =patibility between the general reacter building shell and the area around the cpening.

5 1.1.5 5 2 Icads: The primary, secondary, and ther=al loads that are considered in the design of the cpenings are:

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w e. a .#.#a. . p.3' .g.

. . ~. .~

,g... o. v e .# . k. a. .ek a 7 7

\

3. . , .

OM'8 t

I.

e i

i 4

l I

i

-l Ecvever, it gives an assurance Of the correctness of the assu=2d stress I pattern caused by the prestressing around the Opening.

11. Loads =entioned under e, ab:ve:

With the help of Reference (a), in 51.1551, stress resultants around i the large ope

.ing are fcund for varicus loading cases. Cc=parisc: cf ce results found frc= this reference with the results of a flat plate of veifor= thickness with a circular hole sh0vs the effect of the cylindrical curvature c stress concentraticas around the opening.

Corrections are applied to the stress resultants calculated above to account for the thickening on the cutside face around the large Opening.

These effects are considered using a separate axisy==etric finite ele-

ent ec=puter analysis for both a flat plate and a dc=e with anticipated thickening on the outside face. This finite ele =ent ec=puter progra

4 handles axis 3=ctric and notaxisy==etric leads. The ec=puter result gives six cc=penents of stresses: three ec=ponents for ner=al stress and three co=ponents for shear stress. Inis finite ele =ent ec=puter progra= also is used to predict the effect of concentratics of hoop tendons at the tcp and bctto: cf tme opening.

The other finite element progra= =entioned is the latest one developed *

, at the University of California, Berkeley, which can analyze axis 3=etrie

structures with nonexisy==etric loadings.

\

A constant review is =aintained of the develop =ents of new techniques for analysis and design of large openings. ,

i Based on the past experience of the analysis and design of large openings i (equip =ent hatch openings), the folleving results are of interest:

(1) Tne governing design ec dition for the sides of the opening at the cutside edge of the opening is the accident condition.

(2) Under the condition =entioned in (1).

4 (2.1) Apprcxi=ately 60 percent of the total bonded reinforcing steel needed at the edge of the opening at outside face

, is a result of the ther=al load.

F

-(2.2) Excluding ther=al load, the re=aining stress (equivalent to apprcr*"* ely hO percent of the total lea" *-'"S* g i

! ^=~=' ' cad) at the edge of the outside face is the

contribution of the fciloving stress resultants

_(a) Stresses resulting frc= ce=brane ferces, 1 :luding  :

the effect of thickening, cc: tribute approx 1:ately

=inus 35 percent (cinus lh percent of total).

T

.(b) Stresses resultiDS frc. the =c=ents caused by thicken-ing c 'che cutside face cc: tribute apprcxi=ately 150

,h percent-(60 percent of total).

N._./

5-35 00M9

m (c) Stresses resulting fro: =e=brane force and =orents

\

caused by the effect of cylindrical curvature con-tribute apprerd-ately minus 15 percent (c percent cf total).

In Order t^ "d's

• e the effect of tensile stresses at the outside face and to distribute the concentration Of raiial fer:es (relative to reactor building) exerted by hoop tendons in E =cre unifor:

=anner, vertical tendens in the inside rov vere given a reverse curvature (they are deflected cutward as they pass the opening).

This reverse curvature was given in such a way as to reduce the inward acting radial f:rces (due to hoop tendens) at the top and bottc= cf the opening and to produce inward acting forces on the sides of the opening. It should be noted that n: inward radial force acts on the sides because of the absence of hoop tendons.

12. Ioads centioned under 7 above.

These leads are predicted frc= axisy==etric finite ele =ent ec=puter analysis carried out for dead load, prestress lead and internal pressure load.

Thermal loads vill be established from the co=puted ther=al gradient at the large opening.

The te=perature variation through the concrete vall creates a stress

) condition like one caused by a =o=ent, constant in all directions ss/ cn the centinuous cylindrical or dc=ed surface. However, at any discontinuity, such as an opening, stress concentrations occur.

Using the center of the opening as the reference point to relate the directions of =c=ents, the raiial =c=ent is zero at the edge of the opening, there being no resistance against radial rotation.

The hoop =c=ent is highly increased, the outside fiber being forced to take the shape cf a larger circle, while the inside fiber takes the shape of a smaller circle. Away frc= the edge of the openin6 both =crents gradually reach the constant value on the undistributed portion of the cylinder.

51.1.553 Miscellaneous considerations: creep and thrinkage are not considered in the analysis. However, co=patibility of strain between general vessel shell and the area around the opening is =aintained by thickening the concrete around the large opening.

I; rral shear forces (relative to the Opening) are =t ified to acecunt fer the effect cf twisting moments as sh0Vn in Reference (a) in 51.155.1. These

=odified shear forces, called Kirsch ff's shear fc. :es, are resisted ~cy

'hcrizontal vall ties.

The biaxial cracking changes the stress distribution that is predicted when assuming uncracked concrete. When the cracking is at the outside face, ther=al conents are reduced, ie, self-relieving. For proper control cf crack vidth and spacing, well distributed reinforcing steel is provided in hoop and radial

/"'N direction, ks_-)

• r i

/~-

5-30

f 3

,i In the factored Icei case of 1 5 P i 1.0 Tg (accident te=perature) the cracked concrete with highly strained tensile reinforce =ent constitute a shell with stiffness decreased but still constant in all directicas.

In order to control the increased hoop ther=al = =ent around the opening, the hoop reinferce=ent should produce strength about twice that of the radial one.

In the case of accident te=perature codbined with lov internal pressure, i very small er no tension develops en the outside, so the thermal strains vill be built up without the relieving effect of the cracks. Ecvever, as I-has already been stated elsewhere, the liner plate vill reach its yield

stress, and so vill the concrete at the inside corner of the penetration, i thereby relieving once again the very high stresse;, but still carrying i the high moment in the state or redistributed stresses.

4 The pattern of =e=brane stresses at design acciden'. loading is not expected to be significantly differen* *-+- +he pattern of medbrane stresses during the pressure test, since the =e=brane stresses due to pressure are far =cre siEnificant than those due to te=perature.

5 1.1 5.6 Seis=ic Analysis i

A co=plete dynamic analysis is perfor=ed on the reactor buildings to determine their behavior du inE an earthquake. The analysis is accc=plished in five (5) steps. The first step consists of reducing a typical reacter building into 1 - a =athe=atical model in ter=s of lu= ped = asses and stiffness coefficients.

l 5 The second step is to obtain the natural f-'equencies and mode chapes of the

=odel. The third step is cn evaluation process to deter =ine tha proper values of dn= ping. The fourth step deter =ines resulting internal forces on the typical reactor building using the spectru= response curve of the earthquake.

The fifth step is for internal equip =ent located at different levels and pro-i vides a description of the earthquake environ =ent.

In building the =athe=atical model, the locatiens for lu= ped masses are chosen at floor levels and points considered of critical interest. Between *

! = ass points the structural properties are reduced to unifor= segments of

cross-sectional area and =ocents of inertia. In all applicable cases effective shear areas are co=puted. The foundations of the reactor building ,

are each examined and represented in ter=s of spring constants for rotation and translation. These springs are entered at the base of the model. 'w'ith this infor=ation a c0=puterined analysis is used to for= the stiffness,= atrix, f.K),ofthestructure.

, Themassesarearrangedintoa= ass = atrix,[M].

i The nstural frequencies and = ode shapes are cbtained by solution of the equation belev:

i 2

M X r

lE][X]=w .'J I

i

-Various techniques are available for the solution of the equation. At present, l the =ethod e= ployed is tri-diagonalination by successive rotations. The = ode shapes are plotted and exa=ined to de+.er=ine how the structure is vibrating.

,s All modes are considered with frequencies less than 35 cycles per second.

! k 001.01 5- 37

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5 1.2.1 Tvpes of penetrations a _n p.. . ... . . - .4 . ~ . e.n _  %. ,. - e e..m. e . .,. . e .e.d- .e a , ' e ak *..* e* * , ". e..' a. d. a s s e. u._.* e .e

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See:1on III, fer Class 3 vessels.

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I a. Type 1 - High-voltage power, !.16-) velts and 6900 volts.

s

b. Type 2 - power and control, 600 volts and below.

1 5-9 003.02 w- , .-,-.ew -r.rsw,y---e,2w,--.--,....,y-,w-.,-o py - er. g -r +, e q -- a + n v.-w:- g gw,-. we v. --g y-. g -p sw g.gr p y r y v r - - iy--T' 'M '-

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~

c. Type 3 - Instrumentation, thermocouple leads, coaxial, and other

[ ') special wires.

V Type 1 penetrations are high-voltage insulated copper conductors. These insulated conductors are passed through the header plates and potted to effect a pressure seal. Mechanical splices within the potting compound provide gas stops. High-voltage insulating bushings and seals may also be used to provide the barrier.

Type 2 penetrations are single or multiconductor insulated cable. This cable is connected between the header plate assembly at each end of the penetration canister. Hermetically sealed connectors or potted pressure seals similar to Type 1 are used.

Type 3 penetrations are the same as Type 2 except the conductors are thermo-couple material, ceaxial cable, or special wires. The sealing methods are the same as for Type 2 penetrations.

5.1.2.1.2 Piping Penetrations Single barrier piping penetrations are provided for all piping passing through the reactor building exterior wall. The closure of the pipe to the steel liner is accomplished using forged flued head fittings welded to the 26 pipe and to the liner reinforcement.

In the case of piping carrying hot fluid, the pipe is insulated and cooling 26 may be required to reduce the concrete temperature. Figures 5-2, 5-2A and 5-2B show typical pipe penetration t semblies.

'The anchorages of all the penetratiot to the reactor building wcil are designed as Class I structures to res. all forces and moments caused by a postulated pipe rupture. The desig: nditions include the maximum pipe reactions and pipe rupture forces.

The following design criteria for typica;

• ping penetrations are used to insure the integrity of the liner penetratz s junction at the piping:

1. The penetration assembly and the ass mbly welds and welds to the liner are full penetration welds. The assembly is anchored into the 9011 concrete and designed to secoccodate all forces and aoments due to pipe rupture and th al expansion.

2. The design criteria are such that the pip. cenetration is the strongest point in the system when a pipe brear. is postulated.

Pipe stops, increased pipe thickness or other means are used to attain this. In addition, part of this criteria is that the operation of clostre valves is not impaired by any postulated pipe break.

5.1.2.1.3 Equipment and Personnel Access Hatches An equipment hatch 13 ft-6 in. in diameter is provided as shnwn on Figure 5-3. It is fabricated from welded steel and furnished with a double gasketed

S .

5-39 Amendment No. 2l

()();j){g 4/74 p ,w, y ym +- - ,1

v. ve, ,n,r-- - e, -- - - - , - . ,

O flange and bolted dished dcor. Equip =ent up to and including the sine of the reactor vessel 0-rin6 seal can be transferred into and out of the reac-tor building through this hatch.

Two personnel locks are provided. One of these is for emergency egress only.

Each personnel lock is a double door, velded steel asse=bly. A qu'.ek-acting type, equalining valve connects the personnel lock with the interior and exterior of the reactor building for the purposes of equalizing pressure in the two syste=s when entering or leaving the reacter building.

The two doors in each pers:nnel lock are interlocked to prevent both being opened simultaneously and to insure that one door is co=pletely closed be-fore the opposite door can be opened. Remote indicating lights and annun-ciators situated in the control roc indicate the door operational status.

Provision is cade to per=it bypassing the door interlocking syste= to allow doors to be left open during plant cold shutdown. Each door lock hinge is designed to be capable of independent, three-dicensional adjustment to assist proper seating. An energency liShting and co=nunication system operating from an external e=ergency supply is provided in the lock interior.

5 1.2.1.k Special Penetrations

a. Fuel Transfer Penetration (Figure 5-2): A fuel transfer penetra-tion is provided for fael movement between the refueling transfer canal in the reactor building and the spent fuel pit. The pene-tration consists of c. 36 inch stainless steel pipe installed 3 26 '

inside a 48 inch pipe. The inner pipe acts as the transfer tube d and is fitted with a double gasketed blind fitnge in the refuel-ing canal and e. standard gate valve in the spent fuel pit. This arrangement prevents leakage through the transfer tube in the event of an accident. The outer pipe connects to the reactor ,

building liner.

b. Reactor Building Supply and Exhaust Pure Ducts: The ventilation system purge duct is equipped with two tight seating valves to be used for isolation purposes. The valves are re=otely operated for reactor building purging as described in 9.12.

5 1.2.2 Design of Penetrations 5 1.2.2.1 Design Criteria Penetrations conform to the applicable sections of ASA N6.2-1965, " Safety Standard for the Design, Fabrication, and Maintenance cf Steel Containnent Structures for Stationary nuclear Power Reactors." All personnel locks and any portion of the equipment access door extending beyond the concrete shell confer = in all respects to the _ equirements of ASME Section III, :;uelear Vessel: Code.

Each line which penetrates the reactor building exterior vall and contains high-pressure or high-temperature fluids (steam and feedvater) vill be

[]

v 5-k0 Amendment No. 26 4/74 00 O&

. - - .. _ = - . ~ - . ~ = ~ - - _ _ - _ - -

C l i

I i,

L 4

1 l anchored at the penetration of the reacter building wall. This restricts i

pipe whipping associated with fracture of a line ecstaining high internal j energy and thereby prevents da= age to the penetration and breeching of the  ;

reactor building. ,

2 4

Parther protection of each line, necessary to preclude pipe rapture between penetration and first valve, is accomplished by shortening the exposed length of 3.tpe and installing the first valve as close as possible to the

reacter built'ing external vall, dependent upon valve operating and =aintenance clearances. Criteria which apply to the provision of autc=atic and =anual j isolation valves in the penetration lines are contained in 5 1 5 5 1.2.2.2 Code of High-Temperature Penetrations The =ain high-te=perature piping consists of two penetrations fcr itedvater and two penetrations for =ain stea= vhich have a =axi=u= operating ie=pera-ture range between k50 F and 570 F. Thermal insulation er separate

colant circulation is provided at the penetrations to restrict =axi=u: terrerature

, rise in the concrete to 150 F.

For the condition of less of penetration coclant circulation, the =ax1=u: ,

steady state te=perature in the concrete vill be 300 F at the penetratic:

surface and decrease to 120 F at a =axi=u= depth of k8 inches in the reactor f

building vall. Actual peak te=peratures in the penetrations resulting fro =

, hours. A =aximu=

accidentconditionsareexpectedtosubsidewithin(p/vithoutappreciable temperature of 390 F =ay be tolerated for 120 days N deterioration of the concrete.

, The basis for limiting strains in the penetration steel is the A92 Boiler ,

and Pressure Vessel Code fer Itaclear Vessels,Section III, Article k,19c5, and therefore, the penetration structural and leah tightness integrity is maintained. Iccal heating of the concrete i=nediately around the penetra-i tien develops co.pressive stress in the concrete adjacent to the penetration

and a negligible arount of tensile stress over a large area. The =ild steel reinfercing added around penetrations distributes local ec=pressive stresses i

for everall structural integrity.

5 1.2.2 3 Penetration Materials 1

1 The =aterials for penetrations including the personnel locks and equip =ent

, hatch together with the =echanical and electrical penetrations are carbon s

steel and confor= vith the require =ents of the ADE Itaclear Vessel Code.

l As required by the Itaclear vessel Code, the penetration =aterials =eet the necessary Charpy V-notch 1 pact values at a te=perature 30 F belev the icvest service tenperature.

a. Piping Penetration Materials: luterials srtafications are listed below:

(1) Davis, Harold S., "Ther::a1 Considerations in Design of Concrete Shields,"

ASCE Proceedings, Septe=ber 1956 5-kl 01 @

r

,_7--- .--gg.,,,-,--.v.,,-- ,. , .-..w,,--.y--r- ,mu-w.e e- g- ---ie r - --- we-.*. -r-e+ -n--- =

s

\

s_ / Piping Penetratien Material Specification Penetration Sleeve Welding P-No 1, ASME

' Code, Se: III, Tatle N b21 Penetration Reinferting Rings ASI" - A-516 Penetratien Sleeve Eeinforcing AS24 - A-516 Ear Anchoring Rings and Plates AS24 - A-516 Made to AS24 - A-300 Eclled Shapes AF24 - A-36

b. Electrical Penetratien Materials #

The penetration sleeves to acec==odate the electrical penetration asse=bly canisters are 12 inch, Schedule 80 carbon steel pipe, except where otherwise noted.

c. Access Penetratien Materials The equip =ent and persennel access hatch =aterials are as fellows:

Access Penetratien Material Material Specification b[N Equipment Hatch Insert Equip =ent Hatch Flanges All AS24 - A-516 Made to ASTM - A-300 Equipment Hatch Head Personnel Locks 5.1.2.2.L Double Wall Pipe Penetraticns Bechtel Topical Report BN-TOP-1, Revision 1. Nove=ber 1,1972, " Testing Criteria for Integratr.d Leakage Rate Testing of Pri=ary Contain=ent Structures for Nuclear Pour.r Plants", includes the results of containment integrated leakage rate tests of the Palisades, Point Beach, Turkey Point and other reactor buildings.

These test results substantiate the integrity of the contain=ent structures and de=enstrate that a design leak rate of 0.1% per day can be readily achieved without redundant, or double vall, pipe penetration caps.

'6 The reactor plants listed in BN-TOP-1 used the velded penetration seal cap design. The Midland Plant penetration design nov incorporates the i= proved -

forged flued head fittings instead of the velded seal caps. This design has the following advantages:

(1) The flued head fitting requires fever velds than the seal cap.

(2) The flued head fitting per=its in-service inspection in accordance with

'k',,)N ASME Boiler and Pressure Vessel Code,Section XI.

In view of the above, a second penetration seal on the outside of the containment will not be added and provisions for thstr future addition vill no longer be considered as a design criteria.

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Tha felleving codes of practice are used Ic establish standards cf construe-tien procedure

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a 5 h2a O b, ' <Ar#n.=ent Vi No. 2 5/25/69

ACI 3L7 - Rece=ended Practice for cen: rete Fra=everk ,

ACI 635 - Re::=ende Prac'. ice for H:t Weather Ccs:reting d

ACI 613 - Recc= ended Practice for Selecting Pr pertions fer con rete ACI 611 - Re:0= ended Practice for Measuring, Mixing and Placing Concrete ACI 315 - Manual of Standard Practice for petailing Reinfereed C::: rete i Structures l 1

Part UW - Require =ents for Unfired Pressure Vessels Fabricated by Welding, of Section VIII of the ASMI Sciler and Pressure Vessel Code i

AISC - Steel Manual, C0de of .etandard Practice s

ACI - Inspection Manual AWS - Code for Welding in Building Construction (D 1.0-66)  ;

AWS - Re c= ended Practices for Welding Reinforcing Steel, Metal Inserts and Connections in Reinforced Concrete Construction (D 12.1-61)

In every instance the et.,nstruction procedure for the reacter building equals er exceeds the rec 0=endaticas set forth in the fereg:ing publicatiens.

T The extent to which each detailed process exceeds standard require =ents cannet be described without incorporating all applicable job specifications and anticipating hypothetical construction proble=s and coniitiens.

5133 Quality C= trol Proca:

A tor =al quality control crganization and reperting syste= is e= ployed to assure that critical structures are built in accordance with the specifica-tions. This is generally explained in Appendix l-3.

5133 Construction Materials Inspection and Installation

.WLterials used in the reacter building include concrete, concrete =aterials, reinfercing steel, prestressing syste: =aterials, and staa'

• a- plate.

Tne user inspection and testing cf each =aterial.is as fc11 vs:

5.1 3 3 1 Concrete Materials

a. Ce=ent In addition to the tests required cf the ce=ent manufacturers, the fol10 Vine ; user tests are perfc=ed: -

ASIM C-11L - Che=ical Analysis AS31 C-ll5 - Fiueness of Portland Ce=ent s

5-43 00'C8

ASm C-151 - Autoclave Expansic:

AC4 C-191 - Time Of Set ASW C-lC9 - Oc pressive Strength 27 l Deleted The purpose Of the above tests is to ascertain eccfc.~ance dth AS E Specification C-150. In addi 100, Tests ASW C-191 r d Ac4 C-lo9 are repeated periodically during eenstructie: to check stcrage environmental effects on ee=ent charactcristics. Tne tests supple =ent visual inspection Of naterial storage procedures.

b. Water Reducing Acents A cenerete testing laboratory is engaged to perfom the necessary strength and shrinkage tests c: various water reducing agents to establish the particular additive with the =ost desirable charac-teristice for this application.

i c. Agcregates User tests of concrete aggregate include the following:

T Test Results To 3e

s Asm No. I;a=e Basis For Achieved C-131 Ics Angeles Abrasion ASIM Spe: C-33 To confc= vith Specification C-lk2 Clay In=ps ASTM Spe: C-33 TO 00nfo= vith Specification C-ll7 Material Finer Tnan As m spee 0-33 To confc.= with No. 200 Sieve Specification C-37' Mortar A king AST Spe: C .13 Tc conf 0= vith Properties Specification C LO Crganic Inpurities As m Spe: C-33 Tc cenfe= with Specificatic:

C-289 potential .t.es:tivity ASm Spe: C 33 C: ectf0= vith (Chemical) Specification C-13o Sieve Analysis AST Spec C-33 To ecnfc= with Specifica ic:

C-83 Soundness ASIM Spe: C-33 To confc= vith Specification

\

v) 1 5-*i n tg , .f.g J Amendment No. 27

.8/74

.-_ .. . - , , _._ - - = ._ _ . - - - - ..

i 4

Test Results To Be ASTM No. Name Basis For Achieved C-127 Specific Gravity and ASTM Spec C-33 FEx Design Absorption Calculations C-128 Specific Gravity and ASTM Spec C-33 Mix Design Absorption Calculations C-295 Petrographic ' ASTM Spec C-33 To conform with Specification In addition to the foregoing initial user's tests, a daily inspec-tion control program is carried on during construction to ascertain consistency in potentially variable characteristics such a3 grada-tion and organic content.

d. Water

Water used in concrete mixing will be sampled and analyzed by a qualified testing laboratory to assure conformance with specifica- F tions.

5.1.3.3.2 Concrete

a. Design Mix g

s ,) Design mixes and the associated tests are provided by a qualified concrete testing laboratory. The design of mixes is in accor-dance with ACI-613 to obtain material proportions for the specified concrete. During construction the field inspection personnel make any minor modifications that may be necessitated by varia-tions in aggregate gradation or moisture content.

i

b. -Compressive Strength Concrete strength, slump, and temperature inspections are performed.

The purpose of the test and inspection is to ascertain conformance to specifications. The basis for the proposed inspection proce-dure is ACI Manual of Concrete Inspection with upgraded modifica-tions to meet the core stringent requirements of this application.

5.1.3.3.3 Reinforcing Steel

a. Material All reinforcing. steel is tested by the supplier whose tests are-performed in accordance with ASTM requirements. A minimum of one tension and one bend test per heat or mill shipment, which ever

-25 is less, is required for'each diameter bar No. 11 or smaller.

A minimum of one tension test per heat or mill shipment, which ever is less, is required for No.14 or 18 bars. In addition,

/T U

00" 10 5-45 Amendment No. 25 L 2/74-

the field perfor=s st=ilar user tests en one sa=ple fro = each 25 g 50 tons of =a:erial. F.igh strength bars are clearly identified g ,/ prior :o ship =ent to prevent any possibility of mix-up vith lower strenF:h reinforcing bars,

b. Mechanical Solices The CADWILD inspection program is detailed in Appendix 5-C.

29

c. Fabrication Visual inspection of fabricated reinforce =ent is performed to ascertain di=ensional confer =ance with specifications and drawings.

d. Place =ent Visual Inspection of in-place reinforcenent is perfor=ed by the placing inspector to assure dimensional and location cenfor=ance with drawings and specifica: ions.

5.1.3.3.4 Prestress Syste=

a. Wires Sa=pling and tes:ing of the tendon =sterial used in cens:ruction g confor= to ASTM Standard A-421 or ASTM A-416. The following ,

procedure is used:

)

1. Buttenhead rupture test fre= each reel of wire is =ade.

2. Each sice of wire fre= each =111 hea: and all strands from each =anufactured reel that is shipped :o the site shall be assigned an individual lot n==ber 1:d tagged in such a manner that each such lot can be accurately identified. All unidentified prestressing steel or anchorage assenblies received at the iobsitr: are subject to rejection.

3. Random sa=ples as specified in the ASTM Standards stated above are taken fre= each lot of prestressing steel used in the work.

With each sa=ple of prestressing steel wire or strand that is te:ted, there is submitted a certificate stating the manu-facturer's =ini-" guaranteed ulti= ate tensile strength of ,

the sample tested. Stress-strain curves are plotted and the  ;

yield and tensile streng:h verified. The anchorages develop the -i-i-" guaranteed ulti= ate strength of the tendon and the

= int =u= elengation of the tendon =aterial as required by the applicable ASIM specification.

Field inspectien insures that there are no visible machanical or =etallurgical notches or pits in :he tendon =aterial. '

75

f. )

(10 ' . i.l. ,

5-46 A=end=ent No. 29 4/75

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b. Installation

! Prestressin6 installation verk is inspected by a qualified inspec-tor. All =easuring equip =ent used for installation is calibrated i and certified by an approved independent testin6 labcratory.

During stressing Operations, records are k2pt by Bechtel for use in co= paring force =easure=ents with elongation for all tendens.

The resultant cross-reference provides a " M -heck en ceasure-t

=ent accuracy. Measure =ent accuracy and rejection allevances are in accordance with ACI 318, Chapter 2o.

1

c. Grease Grease is sa= pled after delivery and sub=it ted to a qualified testing laboratory for che=ical analysis to establish confor=ance

with specifications.

51335 Reactor Building Liner Plate a

a. Steel Plate Steel plate is tested at the =111 in full confor=ance to the applicable ASlM specifications. Certified mill test reports are

,t supplied for review and approval by the desi6: Group in the Project En61neer's office. .

v~g There is no i= pact testing done en the liner plate =aterial. The purpose of i= pact testing is to provide protection against brittle failure. The possibility of a brittle fracture of the liner plate is precluded because at the design accident pressure condition

, there is not any tensile stress anywhere in the liner plate.

4 This is true whether there is instantaneous release of pressure er there is sc=e ti=e lag in te=perature lead application. There-fore, the E te=peratura of the liner plate loses significance.

. b. Fabrication and Installation i Welding inspection confor=s to the quality control inspaction i procedure.

Ii=ensional tolerances are checked by an installation inspector to prevent unanticipated installation defc.~ations.

5.1 3.h Specific Construction Topics

! 5 1 3. k.1 Bonding of Concrete Between Lifts Horicental construction joints are prepared for receiving the next lift by vet sandblasting by cutting with an air-water jet, or by bush ha== erin 6 Surface vet retardant ec= pounds are not used.

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When vet sandblasting is e= ployed, it is 00ntinued until all laitance, coating, stains, debris, and Other foreign materials a-a -a Oved. The surface of the concrete is washed th: roughly to remove all loose material.

When air-vater cutting is used, it is perferre: after initial set has taken place but before the 00n: rete has taken its final set. The surface is cut with a high-pressure air-water jet to re=cve all laitance and :: expose clean, sound aggregate, but n t so as to undercut the edges of the larger particles of aggregate. After cutting, the surface is washed and rinsed as long as there is any trace cf cloudiness of the wash water. Where ne -

essary to remove acet=ulated laitance, coatings, stains, debris, and Other foreign caterial, vet sandblasting is used before placing the next lift, to supplement air-water cutting.

H:rizontal surfaces are vetted and covered with one-quarter inch to one-half inch cf =ortar of the same cement-sand ratio as used in the cenerete, i=nediately before the cenerete is placed.

Vertical joints are sandblasted or bush hattered, cleaned, and wetted before placing concrete.

5 1 3.4.2 Prestressing Sequence The detailed stressing sequence is based on the following general require-cents to =ini=1:e unbalanced loads and differential stresses in the structure.

1

a. Every second tendon is tensioned within a strip extending frcr 20 feet ab:ve the construction opening to one 30 feet belev the botto: cf the ring girder,

b. Every fourth hoop tenden is tensioned fra: the botto: cf the ring girder to the previously stressed tendons,

c. All remaining hoop tendens are tensioned within a strip extending frc: 30 feet above the construction opening to one 50 feet below the bottc= of the ring girder.

d. The dome tendons are fully tensioned using a balanced approach.

e. The construction opening is closed,

f. The remaining tendonc are installed including buttonheads.

g. The remaining lover hocp tendons are stressed fcileving the pat-tern used en the upper hoop tendons,

b. The vertical teniens are tensioned.

1. All remaining tendens are stressed.

The procedure for prestressing is carefully worked out with the post-tensioning vendor.

O V 00' '2 5-ha

i J

() All procedures are subject to approval.

The post-tensioning syste= supplier is provided with prestressing forces, the anticipated cracrete elastic, shrinkage, and creep prestress losses, and the maximum prestress forces for each stage of prestressing. The supplier incor-14 porates all this information along with any steel relaxation, friction, and

anchorage losses to establish the initial jacking force for each sequential operation.

3 Force and stress measurements are made by measuring the elongation of the prestressing steel and comparing it with the force indicated by the jack-dynamometer or pressure gage. The gage indicates tha pressure in the jack within plus or minus two percent. Force-jack pressure gage or dynamometer combinations are calibrated against known precise standards just before application of prestressing forces begins and all calibrations are so certi-fied prior to use. Pressure gages and jacks so calibrated arc always used together.

During stressing, records are kept of elongations as well as pressures obtained. Lif t-off stress readings are taken at the end of each stressing operation to check the actual stress in the tendon. Jack-dynamometer or gage combinations are checked against elongation of the tendons and the cause of any discrepancy exceeding plus or minus five percent of that pre-dicted by calculations (using average load elongation curves) is corrected, and if caused by differences in load elongation from averages so documented.

Calibration of the jack-dynamometer or pressure gage combinations is main-

• tained accurately within above limits.

5.1.4 REACTOR BUILDING INSPECTION, TESTING, AND SURVEILLANCE 5.1.4.1 Tests to Insure Liner Integrity As the structure.is constructed, and after it is complete with liner, con-crete structures, and all electrical and piping penetrations, equipment hatch, and personnel locks in place, the following tests are performed:

a. Construction Tests: These take place during the installation of the reactor building liner.

b. Preoperational Tests: These are performed after the erection

< of the structure is complete but before reactor operation.

5.1.4.1.1 Tests on Liner During Construction Inspection procedures to be employed during construction for the liner seam welds, liner. fastening, and around penetrations consist of one or a com-bination of the following tests: visual inspection, vacuum box soap bubble 23 testing, radiography, magnetic particla, and liquid penetrant testing. Non-destructive examination of the liner plate is based on the applicable sections of the Proposed Standard Code for Concrete Reactor Vessels and Contain=ents (Proposed Section III - Division 2 to the ASME Boiler and Pressure Vessel Code) issued for trial use in April, 1973.

i M

()()

• i lk 5-49 Amendment No. 23 11/73

Te further ensure the integrity of the liner, leak chase channels are installed over the liner seam welds and tested in accordance with subsection f below. In addition, provisions will be incorporated into the leak chase channel system to permit pressurization of the channels during operation of the plant, thus 23 providing a means of testing the integrity of the liner seam welds throughout 7 the life of the plant. Thus, since this leak chase channel system provides l

added assurance of weld integrity and hence an added protection against leakage of radioactivity from the reactor building above that provided by most other nuclear plants, it is not necessary to implement the full requirements of Regulatory Guide 1.19.

a. Visual Inspection of Welds All of the welding is visually examined by a technician responsible for velding quality control. The criteria for workmanship and visual quality of welds are as follows:

J Each weld shall be uniform in width and size throughout its full length. The surfaces of welds shall be free from coarse ripples or grooves, overlaps, and abrupt ridges or valleys. Abrupt changes 23 in section thickness such as undercuts or reinforcements which do not exceed 1/32 inch, and which do not encroach on required section thicknesses are permitted. The surface condition of the finished weld shall be suitable for the proper interpretation of nondestructive examination of the weld. If the surface of the weld requires

[) grinding, care shall be taken to avoid reducing the weld or base 1 \- # material below the required thickness. Peening of welds will not be permitted.

Butt welds shall be of multipass construction, slightly convex, of unifor= height, and have full penetration.

Fillet welds shall be of the specified size, with full throat and

legs of uniform length.

b. Soap Bubble Tests i

l All pressure - retaining liner plate welds that are accessible i shall be vacuu= box soap bubble tested in accordance with Sections CC-5521.3 and IX-3700 cf the proposed Section III-Division 2. In this test a vacuum box containing a window is placed over the area to be tested and is evacuated to produce at least 5 psi pressure differential. Sefore the vacuum box is placed over the test area, 23 a soap solution is applied to the weld and any leaks are indicated by bubbles observed through the window.

Acceptance of the soap bubble test results shall be based on.

Section CC-5535 of the proposed Section III-Division 2.

Where vacuu= box soap bubble testing is not feasible, the welds shall be examined by the magnetic particle or liquid penetrant method.

5-50 )3 Amendment No. 23 11/73

( c. Radiography A

Radiography is used as an aid to quality control. The primary purpose of the line.r and the welds therein is to provide leak tightness integrity to the post-tensioned concrete reactor building.

Structural integrity of the reactor building is provided by the post-tensioned concrete and not by the liner. Random radiography of each welder's work provides structural verification that the welding is or is not under control and is done in accordance with the previously established and qualified procedure. Employing random radiography to inspect each welder's work has been proven by past experience to have a positive psychological effect on improving overall welding worknanship.

Radiographic examination of welds is in accordance with the following sections of the proposed Section III-Division 2:

Radiographic Techniques -

Section IX-3300 Examination Frequency -

Section CC-5521.1 Acceptance Standards -

Section CC-5532 At least one 12 inch spot radiograph is taken in the first 10 feet of welding completed in the flat, vertical, horizontal, and overhead positions by each welder. Welders who have satisfactorily welded the first 10 feet of weld shall have one 12 inch long radiograph made of each subsequent 50 feet of weld which he produces.

V 23 Where radiography is not feasible or where the weld is located in ,

areas which will be inaccessible after construction, the welds shall be m mined by the magnetic particle method. The liquid penetrant method shall be used to examine welds where nether radiographic nor magnetic particle examination is feasible.

d. Magnetic Particle (MPE)

Magnetic particle testing is used to examine liner seam welds in areas where radiography and/or soap bubble test methods are not feasible. A minin.um of 10% of all non-radiographable < alds or 100%

of all areas not accessible to soap bubble tests shall be examined as the welding is performed. MPE techniques are specified in section IX-3500 and acceptance standards are.in accordance with sectior CC-5533 of the proposed Section III-Division 2.

Where magnetic particle examination is not feasible, welds shall be examined by the liquid penetrant method.

e. Liquid Penetrant (LPE)

Liquid penetrant inspection will be used to closely examine welds

. Judged to be of questionable quality on the bases of the initial visual inspectiin, and to confirm the complete removal of all defects from areas whL h have been prepared for repair welding. Liquid

(])

'/ penetrant testing is also used to examine liner seam welds where neither radiographic nor magnetic particle examination is feasible.

0006 5-51 Amendment No. 23 11/73

Ihe extent of weld exa=ination in this latter case is the same as

,-s for the magnetic particle method. LPE techniques are specified in section IX-3600 and acceptance standards are in accordance with (s/ ) section CC-5534 of the proposed Section III-Division 2.

23 f. Leak Chase Channel Inspection After the liner plate seam welds have been successfully exa=ined through 1) visual inspection, 2) soap bubble vacuum box tests, and

3) either random radiography or alternately MPE or LPE examination, the seam welds are covered with a leak chase test channel.

29 After attachment of the channel to the liner plate and prior to em-bedding the leak chase channel in concrete, the leak chase cha nnels are pressurized to a minimu= pressure of 80 psi in order to check the integrity of the channel-to-liner plate welds. A soap bubble solution is coated over the attachment welds; any bubble appearing within 20 seconds is cause for repair of that portion of the weld. The test 23 pressure is monitored by valving off of the air supply for at least 15 minutes; any pressure decay noted during this period is further cause for rejection.

In lieu of the tests specified above, the following precedure r:y be authorized. The leak chase channels are pressurized to a mini =um 29 of 80 psi as before. I# any indicated loss of test pressure occurs within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, as evidenced by a test gauge, the channel-to-liner welds shall be subjected to the soap bubble testing.

(S13 Any leakage detected shall be repaired and the channels shall be

( ,j retested.

5.1.4.1.2 Preoperational Integrated Leak Test The design leak rate is not more than 0.1 percent by volune of the contained 14 l atmosphere in 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> at 67 psig. It has been demonstrated that, with good quality control during erection, this is a reasonable requirement.

The basis of the leak rate test is the Absolute Method as specified in Bechtel 26 Topical Report BN-TOP-1, Revision 1, November 1, 1972, " Testing Criteria for Integrated Leakage Rate Testing of Primary Containment Structures for Nuclear Power Plants." During the performance of this test, the liner plate leak chase channels 23 are open to the reactor building atmosphere in order to ensure that the test pressure is applied directly to the liner seam welds.

l The integrated leak test is conducted in accordance with the requirements of 10 CFR 50, Appendix J, " Reactor Containment Leakage Testing for Water Cooled Power Reactors," after the inspection and testing of welded joints, penetra tions, and mechanical closures; co=pletion of repair measures for minimizing leakages; and co=pleti n of any required containment structure pressure tests for strength.

26 Because the reactor building is a thick walled concrete structure, short-ters

temperature or meteorological variations should not have any appreciable effect on

! the reactor building ambient temperature and pressure. The duration of the inte-grated leak rate test is determined by the test duration criteria established in Section 2.0 of BN-TOP-1.

t  !

3 y, Amendment No. 29 5-51a ()($8 ; 5 # 4/75

~. . . .

26 DELETED a

5 1.h.2 strength Test 14l A pressure test is =ade en the ec=pleted building using air at 80 psig.

Tnis pressure is =aintained on the building fer a period cf One hour. During this test, deflection =easu e=ents and observations a-a "da o reasure the response of the structure as described in Appendix 5-E.

1<

5 1.k.3 In-service Tenden surveinance progra:

1 The in-service surveillance progra= for the reacter building tenden syste=

consists cf evaluating the tenden syste= perfor=ance and the corosic prcte -

tion syste= perfomance. Parther, the reactor building structure is a passive type syster where =echanical operational failu es are nonexistent, thus Only requiring that the syste= re=ain at status quo and available to perfc= its function in the unlikely event that it vill be required. It is the intent of the surveillance propa=s to provide sufficient in-service histcrical evidence necessary to maintain the ecnfidence that the integrity cf the reactor building is being preserved. To acc0=plish the surveillance pre-gra=, the following quantity of tendons are =ade available for inspection and lift-off readings:

Hori ontal - Three 2h0 degree tendons ec=prising two ec=plete e hoop syste=s.

Vertical -

Tnree tendons spaced apprcximately 120 dep ees apart.

Doce -

Three tendens spaced approx 1=ately 120 de gees apart.

The surveillance progrs= for structural intepity and c0rosien protectien l[ consists Of the felleving operaticns being perfor:ed during each inspection:

a. 11ft-eff readings vill be taken for all nine tendens.

b. One tenden of each directional group vill be relaxed and th ee l vires re=0ved as sa=ples for inspection.

c. After the inspection, the tendons vill be retensioned to the stress level reasured at the lift-Off reading, and then checked

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d. Should the inspection reveal any significant ecrosion (pitting, j er less of area), further inspection Of the cther two sets vill i be =ade to deter =ine the extent cf the cerosion and its signifi-cance to the lead-carrying capacity of the structu e. Sa_ples of corroded wire vill be tested te failu e to evaluate the effects .

of any cerrosien.

N

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I 0008 5-52 A=end=ent No. 26 4/74

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' D A conservative testing frequency will be established at the operatien license stage and nothing in the design of the reactor building precludes the testing cf tendons or pneumatic testing of the reactor building at any time in its life.

I 5.1.k.k leakage Monitoring Syste:

No continuous leakage =0nitoring system is provided.

The barrier to leakage frc= the reactor building is the 0.e-quarter inch steel liner plate. All penetrations are continuously v'.;ced to the liner plate before the concrete in which they are e= bedded ir placed. These pene-trations, shown en Figure 5-2 becc=e an integral part of the liner and are so designed, installed and tested.

The steel liner plate is securely attached to the prestressed concrete reactor building and is an integral part of this structure. Thorcush centrol is maintained over the quality of all =aterials and work =anship during all stages of fabrication and erection of the liner plate and penetrations and 1

during construction of the entire reactor building.

The co=prehensive progra= for preoperational testing, inspection and post-cperation surveillance is described in detail in 5 1.4 and is su=marized in the following paragraphs.

During construction, the entire length of eve y sea = veld in the liner plate

.' is leak tested. Individual penetration ase " as are shop tested. Welded

- connections between penetration asse=blies and the liner plate are individ-ually leak tested after installation.

Folleving co=pletion of construction, the reactor building is tested at 115 percent of the design pressure to prove structural integrity. The initial

, leak rate testing of the entire reactor building, including penetrations, is conducted at 100 percent of the design pressure and at successively lower pressures to de=cnstrate vapor tightness and also to establish a reference for periodic leak testing for the life of the station. Multiple and redun-

dant systems based on different engineering principles are provided as described in Section 6, " Engineered Safeguards," to provide a very high degree assurance that the maxi =u= pressure and te=perature associated with the acci-dent conditions vill never be exceeded and that the vapor barrier of the actor building vill never be jeopardized.

Under a -" operating conditions and under accident ccnditions including the verst loss-of-coolant accident, no. possibility exists that any leakage

. . coald occur cr that. the integrity of the vapor barrie* -~" be viciated in 1

any way that would be significant to the public health and safety cr to that of the station personnel. Adequate ad=inistrative centrols vill be enforced i to =inimize the possibility of human error. Station operators vill be trained and licensed in accordance with regulations. Safety analyses are presented i

in Section 14.

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-a.=

- -- .e a.e y.-^v4 .. dad. s' *be*

. --. .c a.a_ e. aa" - k.e *.a.s ad --

.c ,eak.a ra.

. m. .k. a. .' .i n a. .- ,- a *. a. 4e -- .

e.*.=. .ed - a ~a e d s *. ^..^ ...-^ . .e .* ^ - k.": - -- .ak. e" * - .' a.

..-s .e a - .d by' a*.ho a.. y.. - . "'o.--s,

• d-

"n a ' .' .e ad e. vida.d

. a.4- --^*a.*.4-* .-. #^-

. . . .. ... k. i ^ ' ^ s~ -

ical and nissile shielding, and for access and operating purposes, also

. a . 4 ^ *. - .c.- *.he

.-ov4.d a. ^- ^--e a *.*.~ a.

• a .'_ 4 - a .* * ^ - wk4^" "

, -.. + o s *. ' +.u+ a. .e -. y ^. *.e '.4-a..-

a,,

... .-,e.a.

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oa,--

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-. . . ----.<.e.. . . e.

7 " .e .' .d =.

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. . e~, *.-k a. .

• e_7 a..-a +e".~. a. c ." *.h a. a* .- s k, a..-a. .d s e"k.j . a. .-*.

to a high degree of tenperature control. The outside of the liner is protected by the reactor building concrete shell v'ich n is exceptionally

.w a. s* ..e.* -= -- *. +^ ev a'-.' vea+ ha.- ao - - d.*..#^*=.m. . 70.- +ha.sa . . . -a. a s ^ - s , .--..- a. *.-' a. adaqua ,"

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^ 4

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2. p.

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g- s 2. personnel Access Lock

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3 Main Stean and Feed-Water Lines N. D . .. -- . . v.

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.- .o*a.d as they are developed, where applicable.

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^

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~ a.

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• e a.x*.a. d -- a dv .' y- k.f- -.'s.e-

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s - a d- "" . .ha..-

described and tabulated in 5 15, " Isolation Systen" and 9 12, " ventilation

.c4, . b. . ...--.n t

v/

Of -[YOI W

5-54

~m

.4 4

1

,(

A 515 IS01ATION SYSTEM

5151 Design Bases 1

i The general design basis governing isolation valve requirements for reactor 1 building fluid penetration is:

l leakage through all penetrations not serving accident-consequence-74-4 ting syste=s is te be ~4 **~4 ed by a double barrier so that no single, credible failure or =alfunction of an active cc. ponent can result in less-of-isolation.

The installed double barriers take the for: cf closed piping syste=s, both inside and outside the reactor building, and various types of isolation valves.

1 I

Reactor building isolatica occurs on a signal cf high pressure in the reacter building. Valves which isolate penetrations that are directly open to the reactor building at=csphere such as the reacter building purge valves and su=p drain valves vill also be closed on a high radiation signal. Develop-

=ent of the instru=entaticn circuits and instru=entation signals is presented in Section 'i.

The isolation syste= closes all penetrations, not required for Operation

of the engineered safeguards syste=, to ~4 *~d'a *he leakage of radioletive

=aterials to the environ =ent. In addition, all isolation valves, upon loss 4 _s cf actuating power, fail closed except those required for engineered safe-guarns.

All remotely operated reactor building isolation valves are provided with control switches and position li=it indicating lights in the control roc =.

5152 syste= Design

, The fluid penetraticas which require isolation after an accident =ay be

classed as follows

i

a. Type I

! Each line connecting directly to the reacter coclant syste has tvc reactor building isolation valves in series. These valves ,

l' =ay be either a check valve and a re= tely cperated valve er two re=otely. cperated valves, depending upon the direction of n0=al i

flow.

b. Type II Each line connecting directly to the reacter building at=osphere has two isolation valves in series. These valves =ay be either i a check valve and a re=0:ely operated valve or two re=0tely operated valves, depending upon the direction of nor=al flow.

O

\%

00 T1 5-55 n..c,... , , , , - . , - - . _ ~ . - . . . . , . . , -

_. .- .- _ , . _ . .. - - _~ . - . . . . - .- . . -

4 1

S I c. Tvre III e

l Each line net directly connected to the reacter coolant syste=

er not open to the reactor building atmosphere has one valve, ,

either a check valve or a remotely operated valve, depending en the direction of nor=al flev.

l d. Tvre IV i

Lines serving engineered safeguards systems have isolation valves which are automatically cperated by Safety Injection Signal or i remotely fro = the control roc =, hence are not autc=atically actuated l by the reactor building isolation signal.

Additionally, there are various arrangements in each of these majer groups.

4 The individual system flow diagrams show the =anner in which each reactor l building isolation valve arrange =ent fits.into its respective syste=, For

. convenience, each different valve arrangement is shown in Table 5-2 and 4

Figure 5-8 of this section. The synbols on this figure are identified on Figure 9-1. The table lists the mode of actuation, the types of valves, and their normal positions. The specific syste= penetrations to which each of these arrangements is applied are also presented. -Each valve is tested periodically during nornal operatien or during shutdown conditions to insure its operability when needed, j l 5'.l.5 3 Penetration Pressurination Syste=s t

Tvc systems are provided for preventing possible leakage through reactor

, building penetrations, the isolation valve seal vater syste= and the air

pressurization system.

s J

a. Isolatien valve seal vater syste= (see-Figure 5-9).

h This syste= pressurizes, with water, the pipe penetrations which j are either connected to the. reactor coclant. system or open to the

reactor building at=csphere, except these for engineered safeguards

i. systems. This pressurization is effected by ' seal vater injection j lines that supply, through a set of valves opened-ty the reactor

building isolation - signal, the de=ineralized water ' contained in a pressurized storage tank. lZhe seal water is fed into the pipe section between the two required isolation valves or between the required two sets of isolatien valve sten packing. The seal water is, injected at a pressure slightly higher than the reacter building design' pressure.

i

- l tng ter= operation following a loss-cf-coolant accident the

assurized storage. tank is =aintained adequately fille vith theup fro

the reactor plant makeup water syste=.

The-following reactor building penetrations are provided with

,  ! automatic seal vater injection:

h.

r-Amendment no. 5 ,

i 11/3/69 5-56  :

00 ,, .,e,,,

• g- em r y e-p-p+,y ea 9 e,qw,te,,,s e- ey -+= e-v --g-, u.y y g - - r m.' 4 'r ^7-a v =v -N ' *

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4

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. Reacter cociant syster d"ain tank vent and discharge.

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between two valves in series such as the tutterfly valves in the

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-s" 11/3/69 5-Sea

c. Liner plate leak chase channel pressurization syste=

26 As discussed in Section 5.1.4.4, no continuous leakage monitoring system is l ,,

'9 provided. The channels over liner plate welds provide a means of testing the integrity of the liner plate sea = welds throughout the life of the plant. 'Ihey are not normally pressurized and are not pressurized after a loss of coolant accident. They may, however, be used to periodically test the containment leak tightness. If the results show that the leakage rate exceeds the design rate, 26 the channels will be pressurized with air from the penetration pressurization system until plant operations permit scheduled repairs.

Since the channels serve no safety function and since no credd.t is taken for the channels in the analysis of accident radioactivity re leases, the channels and their connecting piping are not regarded as engineered safeguard equipment.

O t

I<

(N

\_

5-56b 0042d

~

Amendment No. 29 4/75

.~ . _ _ _ . _ _ _

c. Liner plate leak chase channel pressurization rystem j As discussed in Section 5.1.4.4, no continuous Jeakage monitoring system i

is provided. The channels over liner plate walds are used to check leak tightness prior to placement of the adjacent concrete. They are not normally pressurized and are not pressurized after a loss of coolant accident.

26 They may, however, be used to periodically test the containment leak tightness. If the results show that the leakage rate exceeds the design rate, the channels will be pressurized with air from the penetration pressuriza-4 tion system until plant operations permit scheduled repairs.

j Since the channels serve no safety function and since no credit is taken

for the channels in the analysis of accident radioactivity releases, the

, channels and their com.ecting pipin;; are not regarded as engineered safeguard

equipment.

3 I ,/

i i

1 I

i 4

_O O

5-56b Amendment No. 26 00 ?,"5' 4/74

- . , . - . - . - . . - . . . . . . . _ . . - . . . - . - _ . . . . . . - . - _ . . . . - . .. . n .. ,

v. (,

(

a 4

) Table 5-2 4

. Reactor Building Isolation Valve Arrangements Norrwil j Penetration Flow Valve Valve 2

Number Service Direction Arrangement Type Signal Position

l' Pressurizer ES Closed I. Cample. Lines Out 1 I ES Closed i i 2 .. . Steam Generator 1- Secondary Water

i. Cample -Lines Out 2 III ES Closed

i. i i 3 Component Cooling

} Water Inlet Line In 4 III ES Open 1

! Y 4 Component Cooling i $ Water Outlet Line .Out 2 III ES Open

!- 5,6,7,8 Reactor Building Air

!. Recirculating and I

! Cooling Units Cooling Water Inlet In h IV ES Open  !

'.; O 9, 10, 11, 12: Reactor Building Air g

... Recirculating and 4-

, p; Cooling Units 7;) . Cooling Water Outlet Out 3 IV ES Open j

{-

ES Closed

i. .

t i I

.13, 14 Reactor Building Normal and Emer- ES Closed gency Sump Drains Out 1 II ES Closed I

i 15 Intdown Line to Purification ES Open

{

i Demineralizers Out 6 I EG Closed j EG Open 1

k I

a

l-  :(

i i  !

iL Table 5-2 (Contd)

Normal i

' Penetration- Flow Valve Valve  :

i- Numb < Service Direction Arrangement Type Signal- Positton a

l 16 Reactor Coolant

?- pump Seal Water -- Open i Supply In 7 I dp Throttled  !

[' 17 Reactor Coolant  ;

! Pump Seal ES Open Water Return Out 12 I ES OPn

{. 18 Normal Mikeup to

Reactor Coolant -- Open I

System In 8 I Preocurizer Throttled 4 Icvel

! Y I t g-. 19, 20,.21, 22 Righ-Pressure --

Closed

_ Injection IJne In 9 IV ES Closed ,

! 23 Fuel Transfer --

Closed l Tube 'In/Out 10 II --

Closed j 24, 25 Reactor Building ES Closed

] Spray Line ' In 5 IV --

Closed Q 26, 27 Decay Heat Removal -- Closed Inlet Line In 9 IV E3 Closed f

N 2d Decay Reat Removal RMC Cloacd i Ptunp Suction Line Out 11 I RMC Closed ES Cloned 1

29, 30 Reactor nailding i anergency Sump ES Closed l Recirculation Line Out 2 IV i-i

(  ;

i i

_ . - . . . _ . . _ _ _ . . _ _ - . . - _ . . - ~ . . . _ . . - _ . _ .-- _ . . . . _ _ _ . . . . . _ _ . _ . _ . _ _ . - _. _ _ _ _ . . , . . .

. O O l Table 5-2 (Contd) i  !

Normal Penetration -Flow Valve Valve

-Number Service Direction Arrangement Type Signal Pc.sition i 31 Reactor Coolant 1 System ES Open

Drain Tank Vent Out 1 II ES Open I .

32 Reactor' Coolant System

' Drain Tank ES Open

• Discharge.Line Out 1. II ES Open 1

i

33 Reactor Building .

ES Closed

. , . Purge Inlet In 13 II ES Closed

!' O w-i l 34 Reactor Building ES Closed j' Purge Outlet Out 14 II ES Closed ES Closed 4 35, 36 ' Auxiliary Feed-Water Line .In 15 IV --

Closed 37, 38 - bbin j- g . Feed-Water Line In 15 III --

Open i

i

h. 39,.ho ruin Steam Line Out 16 III High-L.>, Pressure Open or l tj ICS Closed as l ES Appropriate L

{ 41 Dominera11 zed Water Supply to l Pressurizer Quench --

Closed i Tank In 5 II ES Closed i

! '42 . Service Air _

Closed Supply LLne In 5 II ES Closed i  !

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!n) v 5 1.6 DESIGN EVALUATION The reactv. building with the appurtenant engineered safeguards systers pre-vents uncontro21ed release of radioactivity to the envirens during nc: a1 plant operation cud in the event Of a loss of coclant accident. Tne reactor building is designed for the pressure and tenperature resulting fr0= a loss-of-coolant accident, concurrent with other design leads. Tne adequacy cf the reactor building to withstand design pressure and to licit leakage to the design value vill be de=cnstrated by tests at the conclusion of construc-tion."

52 AtJXILIIGY EiJIIDIN3 5 2.1 DESIGN BASES The following cignificant facilities related to plant safety are located in the auxiliary build N

1. New and spent fuel handling, storage and shipnent facilities.

2. Control roO: and related facilities.

3 Radvaste decontanination facilities.

h. Radvaste cherical and volune cc ntrol facilities.

(m)

,/

w 5 Access centrol roc =.

6. Engineered safeguards syster.

7. Reacter building penetration areas.

The. above areas of the building are designed fcr the fclicving loads, if applicable:

1. All norral dead, live, external hydrostatic, vind and seistic loads acting upon the=. The live loads are in accordance with the specified codes in the design criteria, Appendix 5-A.

2. Eigh-velocity wind loads due to tornado and the effect of =issiles generated fran tornado, except fer the 3'n'ST and the steel frece encicsure Over the spent fuel pool.

3 Internal flooding due no pipe rupture and the resulting hydrcstatic lead.

L. Vertical, lateral and stean jet loading resulting frcn rupture cf high-pressure piping.

5 Special requirements to prevent criticality of new and spent fuel bundles.

All areas of the auxiliary building, except the metal siding above the fuel FT handling deck, are designed to Class I standards.

iv  !

5-62 00'31 ^=endrent No. 2 5/28/69

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\'# I The feed-water lines and the main steam lines are enclosed in separate concrete enclosures. These lines are restrained by structural steel franes, anchcred to concrete slabs er valls, to prevent damage to cne ancther cr to any other critical systems due either to a double ended or a slot type cf failure.

5.2.3 NUCLEAR FUEL STORAGE CONSIDERATIONS The spent fuel bundles are stored in rigid stainless steel racks. There are two plates between each pair of bundles, throughout the active length of the fuel bundles. This and the geometry serve to avoid criticality, assuming unberated water in the fuel pool. In additien, the fuel pool is filled with beric acid solution. A storage space fcr fuel bundles has been provided which amounts to 1-2/3 of the reacter core. The valls and floor of the fuel pocl are lined te provide leak tightness. Facilities for inspec-tien of failed fuel bundles have been provided and four spaces are also provided for their storage.

The new fuel bundles are stored in a rigid rack. Prctection against criti-cality in water is net required since the stcrage area has a grating flocr to avoid flooding.

5.3 OTHER PLANT STRUC"'URES 5.3.1 TURBINE EUILDING

(',\.

\- a subdrain system is provided around the turbine building to reduce the groundwater level. In the event of a severe flood cr a failure in the drainage system the groundwater vill relieve into the turbine building.

Critical switchgear and safeguards equipment are located in the auxiliary building thereby protecting against the probable maxinus f1 cod with stage at elevation 632 ft.

5.3.2 SERVICE WATER PUMP STRUCTURES The folleving facilities related to plant safety are located in er near the service water punp structure. They are:

Service water pumps.

Fire water pumps.

Cocling tevers and basins.

Energency storage basin.

30 Pipe lines connecting the emergency cooling water pond and the service water pump structure.

m i \

\s i 5-6h n t )( 1 , ,>, ,^ ,

Anendnent No. 30 9/75

m k,) The service and fire water pumps are located inside reinforced concrete enclosures designed to withstand the loads in Appendix 5-A for Class 1 structures including tornado and maximum probable flood. It is assumed that, in the event of a cooling tower failure due to tornado, earthquake, or other causes, the supply water normally returning through the tower basin to the service pumps will be diverted. For this reason the pu=ps are provided with a direct connection to the pond which is capable of supplying cakeup to the service water system. An emergency supply of water is provided by a depression in the bottom of the pond in the unlikely event of a cooling pond dike failure.

The intake and return pipe lines from the service water pump structure to the emergency cooling water pond shall be prestressed concrete water pipe, embedded steel cylinder type, conforming to American Water Works Associa-tion specification AWWA C-301-72, " Standard for Prestressed Concrete Pressure Pipe, Steel Cylinder Type, for Water and Other Liquids." These lines are classified Seistic Category 1 and will be designed to accom=o-date postulated seismically induced ground motion without loss of function.

The pipe line will be designed to acconcodate seismic ground motion as well as differential motion with the service water pu=p structure. The pipe line vill be provided with sufficient flexibility by the joint design to allow longitudinal movement in either direction along the axis of the pipe and to allow an angular deflection.

The seismic analysis of the pipe line vill be based on the principles con-

[\- /

)30 tained in section 6 of reference 1 below. References 2 and 3 below will be used as guides for the design of the pipe line.

1. BC-TOP-4-A Revision 3, " Seismic Analysis of Structures and Equip-ment for Nuclear Power Plants," Bechtel Power Corporatien, Nov.

1974, (pp. 6-1 through 6-13).

2. N. M. Newmark, and E. Rosenblueth, " Fundamentals of Earthquake Engineering," Prentice Rall (1971).

3. N. M. Newmark, " Earthquake Response Analysis of Reactor Structures,"

Nuclear Engineering and Design, Vol. 20, pp. 303-322 (1972) .

5.3.3 DIESEL GENERATOR BUILDING Two diesel generators are located in the diesel generator building. The building and equipment foundations are designed to withstand the loads in Appendix 5-A for Class I structures including tornado and naximum probable flood. Fire walls are provided between each diesel generating unit.

A 5-65 00.M Amendment 30 9/75 s

s 5.3.4' ADMINISTRATION AND SERVICE BUILDINGS i

Buildings included in this category are designed to withstand the loads in Appendix 5-A for Class II structures.

5.4 RADIATION PROTECTION 5.4.1 RADIATION ZONING AND ACCESS CONTROL The following list identifies the different zones used for the Midland Plant:

Design Dose Rate Zone (cRem/h on a Designation 40/h/k'eek Easis) Description I s 1.0 Uncontrolled, unlimited access.

s 2.5

()

II Controlled, unitnited access.

40 h/ week.

III s 15 Controlled, limited access for routine tasks. 6-40 h/ week.

IV i 100 Controlled, limited access for short periods. 1-6 h/veek.

V > 100 Controlled occupancy for very ,

short periods. Occupancy during emergencies. Normally inacces-sible.

UNCONTROLTID areas are those that can be occupied by plant personnel or visitors on an unlimited time basis with a mini =uc probability of health hazard from radiation exposure.

O("'s tT)0 ; '3.Es 30 5-65a Amend =ent 30 9/75

4 CONTROLED areas are those where higher radiation le"els and/cr radioactive contamination which have a greater probability of radiatien health hazard to individuals can be expected. Nor= ally, only individuals directly invcived in the operation of the plant vill be allowed to enter these areas.

ACCESS 52 areas are those that vill receive radiation dose rates cf less than 100 ne=/h and which can be entered either through open passages cr unlocked docrs. These areas can be entered by all individuals who have passed through the plant access control station.

IRACCESSEE areas are those where dose rates above 100 =Re=/h can be expected.

These a-eas are either blocked off ec=pletely or can be entered only through locked doors.

Centrolled areas are identified by radiation caution si6ns at strategic locations. Restrictions are enfereed by re=0vable cen: rete shielding blocks, l locked doors, chains, etc. Access is supervised frc= the access centr:1 station and the plant control roc =.

In case of e=ergency, personnel vill be able to use escape routes which in-volve the -ini=u= exit time.

5.L.2 RADIATION SEELDING 5.h.2.1 Design Bases i

The basis for the shielding design for nc.,:al plant operation is the " Code of Federal Regulations," Title 10, Chapter 1, Part 20, entitled " Standards for Protection Against Radiation." The exposure of individuals to concen-

! trations of radioactive =aterials in air or water, abcVe the contributions frc= natural background, is limited to values in Appendix 3, of Title 10, Chapter 1, Part 20, of the Code of Federal Regulations.

All areas of the plant are subject to these regulations. The areas are zoned acecrding to their expected occupancy by plant personnel and their designed radiation exposure level under nor=al cperating conditions, i Allowable desi 6: dose rates for all accessible areas of the plant are a j =axi=ue whole body ex vasure of 1.25 Re: per calendar quarter. For all

areas outside the plant, the allowable dose rate is not =cre than 0 5 Re=

, for one calendar year.

i The shielding design is based on the radiation exposure li=1ts set forth l . in :ne Code of Federal Regulations, Title 10, Chapter I, Part 100, " Reactor 1 Site Criteria."

! No individual vill receive =cre than 25 Re: cf whole body exposure er 300 Re: cf thyroid exposure during the course of the accident. This applies l' to percennel who are in=ediately evacuated frc= the plant site and to personnel who re=ain in the central control roc = for up to'10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> and subs:quently leave the plant site.

v) 1 .__ _ . _ , . _ _ _ . _ _ .

i O

%- 2 Allowance is made in the calculation of these doses for short duration trips to critical equipment for minor maintenance.

54.2.1.1 Radiation Exposure of Materials and co:ponents No regulations similar to those established for the protection of individuals exist for materials and co ponents. Materials are selected on the basis that radiation exposure vill not cause significant changes in their physical properties which adversely affect their operation during the design life of the plant. Materials for equipment required to operate under accident conditions are selected on the basis of the additional exposure received.

5 4.2.1.2 General Design Considerations The shielding design considers three conditions:

1. Fall core power operation at 2552 MWt. This also includes shielding requirements for certain off-nor=al conditions such as the release of fission products from leaking fuel ele =ents.

2. Shutdown. This condition deals mainly with the radioactivity from the suberitical reactor core, with radiation from spent fuel bundles during on-site transfer, and with the residual activity in the reactor coolant and neutron-activated materials.

I /~'\ i 3 Accident.

'b' A list of publications and computer progra=s which are used in the design of the radiation shielding is given in 5 4 3 5.h.2.1 3 Specific Design Values The material used for most of the plant shielding is ordinary concrete and concrete block with a bulk density of 144 lb/ft). Only in a very few in-i stances vill steel or water be utilited as primary shielding naterials.

5.4.2.2 General Descriptions and Evaluations 5 4.2.2.1 Reactor Building The reactor building serves two main shielding purposes:

1. During cperation, it shields the surrounding plant structures and yard areas fror radiation originating at the reactor vessel and the p"imary loop co=ponents. Together with additional shielding inside the reactor building, the concrete shell vill reduce radiation levels outside the shell to below 1.0 = Rem /h in those areas which are occupied by personnel either on a permanent or routine basis.

' 2. In the event of an accident, the shielding vill reduce plant and off-site . radiation intensities, emitted directly from released

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