ML20008D766

Chapter 4 to Midland 1 & 2 PSAR, Rcs. Includes Revisions 1-36

ML20008D766
Person / Time
Site:	Midland
Issue date:	01/13/1969
From:	CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
References
NUDOCS 8007300651
Download: ML20008D766 (52)

Text

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TABLE OF COI. TENTS Y

Section Page h EEACTOR C00LAI? SYSTDi h-1 h.1 EESIGN EASES h-1 h.1.1 Fr..-50D'.AI;CE OBJECTIVES h-1 i

h.1.2 EESIGN CHARACTERISTICS h-1 h.1.2.1 Design Pressure E-1 h.1.2.2 Eesign Temperature h-2 h.1.2 3 Beactien wads 4-2 Seismic icans 4-g h.1.2.h

. h.1.2 5 Cyclic Leads 4-2 h.1.2.6 'w'ater Chemistry h-2 h.1 3, EXPECTED O?E".ATING CONDITIONS h-2 5 h-3 ff h.1.h SEFVICE LIEE h.1.h.1 Material Eadiation Darage , h-3 h.1.h.2 Nuclear Unit Operational Tner a1 Cycles h-3 Operating Procedures k-b h.1.h.3 h.1.h.h Quality Manufacture h-5 h.1 5 CODES AI;D CLASSIFICATIONS L-5 h.2 SYSTDI DI.SCRIPTION AND OPERATION h-5 h.2.1 GENERf6 DESCRIPTION h-5 h.2.2  !.'AJOR CC'?OTEI.7 S h-5 h.2.2.1 Reactor Vessel h-5 h.2.2.2 Pressurizer h-6 0 0 uJ*w h7 h.2.2 3 Steam Generator L-o h.2.2.h Peacter Coolant Punps_

U Ee a ct e r Ceci a r.: Pipin- THIS DOCUMENT CONP.lNS "-1^a ulL.2.2.5 POOR QUALITY PAGES j

~ ' ' ' '

a o .

l TAELE OF COI.TE;TS (Contd)

) -

Section rage h.2 3 FRESSURE-RELIEVIIG DEVICES h-10a

-26 k.2.h E; VIRO :s ;TAL PROTICTION h-10a h.2 5 1.'ATERIALS OF CO:;5-'EJCTION h-11 h.2.5.1 General h-11

.h.2 5 2 Pressure Vessels h-11 h.2.6 MAXI:CJM REATIEG AND COOLIIU PJ2ES h-12 h.;.<-

. T_r v. w r .m::

- a h ,. s-h,3 SYSTE4 DESIGN EVAL'JATION h-13 ,,

~h .'3 1 SAF M FAC2 ORS h-13

-h.3 1.1 Pressure Vessel (Includine Stes: Generator) Safety h-lh k.3 1.? Pipin: h-18

g h.3 2 EELIr

CE ON II;TERCONIiECTED SYSTE!S L-18 h.3 3 SYSTEi INIERITY h-10 4.3.k FRESSURE RELIEF h-19

h.3 5 'FJE NDANCY L-19

'h.3 6 SAr m ANALYSIS h-19 h.3 7 OPERATIONAL LDiITS h-20 h.h: ' TESTS AI;D IKSFECTIGUS h-20 h;h.1 .CC:GONE.'? IN-SERVICE IRSFECTION h-20 h . h'.1.1. Feacter Vessel h-21

^

h.h.1.2 ~ Pres suriter h-21 h . k .1 3 ' Sten: Generater- h-21 L.h.1. - Peactor Coclant' Pumps h-21 h.h.1 5 -Fiping h m.

(f[* h . h . _j 6 Iissisilar ::etal and Pepresentative Welds gypA h-22 h 44

^- A endnent No. 26-

. !./ 7/.

TABLE OP CONTEhTS (Contd)

-Section Pa e

.h.h.1 7 Inspection Schedule h-22 h.h.2 PSCTOR COOIki.T SYSTDi TESTS AND INSPECTICNS h-22 4.4.2.1 Reactor Ceolant System Precritical and Hot Leak Test h-22 h.h.2.2 Pressurizinr System'Irecritical Operational Test h-22 h.h.2 3 Relief Syste= Test h-22 h.h.2.h Unit Power Startup Test 4-23

-h.h.2.5 Unit Power Eeat Balance h-23 k.k.2.6- Unit Power Shutdown Test h-23 h.b.3 MATERIAL IRRADIATION SURVEILLANCE h-23 h . 5' . QUALITY CONTROL h-2h h.5 1 GENERAL h-2L

~

h.5 1.1-' Dimensional Inspection h-2h h.5 1.2 Nondestructive Testing "h-2h k.5 1 3 Welding and Heat Treating Processes h-26 i k . 5 1. h' Material Identification h-27

.h.5 2- PHCTOR VESSEL h-27 ll'h;5 3 STEA51 GENERATOR h-27b

'h,5.h- PRESSURIZER h-25 h.5 5 PHCTOR COOLAI.T. PIPIIG h-29 h . 5. 6~ - REACTOR. COOLALT PUMP CASINGS h-28 7~.

L.6  : REFERENCES- h-29 e

t

. xa 00359 -

h-~iii Arendrent Nc. 2 5lIElE ^-

LIST OF TAM FS 3

( At Rear of Section)

I Table No. Title Page L-1 Tabulation of Eeactor Coolant Systen Pressure Settings L-20 li-2 Eeactor Coolant Quality L-30 L-3 Reactor Vessel resign rata h-31 LL Pressurizer resign Data L-31 L-5 Stea Generator resign Iata L-32 L-6 Stean Generator Teedvater Quality L-33 L-7 Eeactor Coolant Punp resign Iuta L-3L L-8 Reactor C clant Piping resign rata L-35 L-9 Transient Cycles L-35 L-10 resign Transient Cycles L-36 L-11 Reactor Coolant Syste: Codes and Classifications L-36 29l4-11a ASME Code Cases 4-36a 4-12 Materials of Construction 4-37 4-13 References for Figure 4 Increase in Transition Terperature Due to Irradiation Effects for A302B Steel 4-39 00360 L-iv /rendnent No. 29 4/75

(N LIST OF FIGURES (At Eear of Section)

Figure No. Title L-1 F.eactor Coolant Syste=

h-2 Reactor Coolant Syste: Arrangement-Elevation h-3 Reactor Coolant System Arrangement-Plan hh Nil-Ductility Transition Temperature Increase Versus Integrated Neutron Exposure for A3023 Steel h: Reactor Vessel h-6 Fressurizer h-7 Steam Generator-h-8 Steam Generator Heating Regions h-9 Steam Generator Heating Surface and Downe::er Level Versus

?over h-10 Steam Generator Temperatures h-11 Reactor Coolant Pump h-12 Predicted NUTT Shift Versus Reactor Vessel Irradiation

> 00 eof.

L-v

(- .

O

'(g h RFACICR CCOLA'~P SYSTEM 4.1 EISIGN IASES The reactor ecolant systen consists of the reactor vessel, coolant pumps, steam generators, pressuriner, and interconnecting piping. The functional relationship between coolant system cetponents is shown in Figure L-1. The coolant systen physical arrangement is shcvn in Figures h-2 and h-3 The reactor ccolant systen cenponents are designed in accordance with the following codes:

1 Piping & Valves - USASI B-317 - February 1968 plus June 1968 errata -

1967, including nuclear cases.

Pump Casing - ASME 3 oiler and Pressure Vessel Code,Section III, Nuclear Vessels, as applicable.

Steam Ginerators - AS!:E Soiler and Pressure Vessel Code,Section III, Nuclear Vessels.

Pressurizer - ASME Boiler and Pressure Vessel Code,Section III, Nuclear Vessels.

Reacter vessel - ASME Boiler and Pressure Vessel Cede,Section III, Nuclear Vessels.

To assist in the review of the system drawin6s, a standard set of symbols and

[] abbreviations has been used and is sutmarized in Figure 9-1.

k.1.1' PEOPy_ANCE OBJECTIVES The reacter coolant syste= is designed to contain and circulate reactor coolant at pressures and flows necessary to transfer the heat generated in the reactor core to the secondary fluid in the steam generators. In additien to serving as a heat transport mediu=, the coolant also serves as a neutron cederator and re-flector, and as a solvent for the soluble boron utilized in chemical shin re-

. activity control.

As the coolant energy and radioactive caterial container, the reacter coolant system is designed to maintain -its integrity under all operating conditions.

While perforcing this function, the system serves the safeguards objective of rininining'the. release to the reactor building of fission products that es-cape the prinary barrier, the . core cladding.

4.1.2 DESIGN CHARACTERISTICS h'.1.2.1 Eesign Pressure The reactor ccolant syster design, operating, and control set point pressures g are111sted in Table b-1. The design pressure allows for operating transient

pressure changes.

" C' The selected design targin considere core ther .21 lag, cool-yv ant transpor, tines and pressure drops,-instrumentation and centrol rc p:nse 00002 h-1 A r. dr 2 nt : : .. u. 5

l

) l 1

i 1

1 characteristics, and systen relief valve characteristic . The design pressures 3

and data for the respective syster components are listed in Tables L-3, L L, h-5, L-7 and L-8. l l

h.1.Ac.ce *s

• recicn 2errer'.ture m

t The design temperature fcr each ccrpenent is selected ateve the maximu ertici-pated coolant temperat,ure in that cctronent under all n:r:21 and transient 1 cad ccnditiens. The design and cperating temperatures of tne respective systen cct-ponents are listed in Tables L-3, L-h, h-5, L-7 and h-S.

h.l.2 3 Eeacticn Loads All ec penents in the reactor coolant systes are supported and intercennected

. so that ri. ring reaction fcrees result in ec bined techanical and therral stresses l in equiptent nc :les and structural walls within established ccde limits. Equip-cent and pipe supports are designed to abscrb piping rupture reacticr leads te ninizine secondar" a accident effects such as pipe actic: <

.nd equip.ent fcandaticn shifting.

I

! h.l.2.L Seismic Loads Eeactor coclant systes ec=ponents are designated as Class I equipment, and are h designed to maintain their functional integrity during earthquake. The basic design Suide for the seismic analysis is the AEC publication TID-7C2h, "!uclear Reacters c-and Earthquake." Equipment will be designed in accordance with Appendix

/M*

I h.l.2.5 Cyclic Loads i

., All components in the reactor coolant systen are designed to withstand tne ef-

) fects of cyclic loads due to reactor syster terperature and pressure chances. ~

I These cyclic 1 cads are introduced by normal unit load transients, reactor trip, l and startup a*.d shutdown operation. Design cycles are shcwn in Table L-9 D r-ing unit startup and shutdown, the rates of terperature and tressure chances ^ ~

are litited.

h.l.2.6 Water Chemistry The water chenistry is selected to provide the necessary tercn ecntent fcr re-activity centrol and to minimize corrosien cf reactcr ecclant syster surfaces.

The reacter ecclant chenistry is discussed in further detail in 9.2.

k.,.-, ry . r.m..:D

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Throughout the load range frc: 15 to 100 percent power, the reactor coolant systen is operated at a constant average temperature. .eactor ecolant syster pressure is controlled to provide sufficient everpressure to maintain adequate core subcocling.

The cinicus cperating pressure is estatliched frc ccre IN rmal analysis. This a.. . .,. .s e. .i s

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c. . c' . k. . .3 '. a. 's . a . *. "a. e cf the vessel materials. The various cc binations of stresses are evaluated and e y.1 yed

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. . e c-r a.-a' i a e- *f*.-^ -. c a '" ~ e e .

The basic determination of vessel c.reraticn frc cold startur and shutdevn to .

full -ressure , and te -erature r c.reration is terforced in a ccrdance with a " Frac-ture Analysis Diagram" as published by Fellini and Punak. bn n' '. *. e_ , e.- a* u e =. b a.2 ' c t .. .b a" ' l "~" c '. .~4 ' # ., ~~

. .. a.. e i + i c.. ". e. , a_ . a .u.~. ( ". . . ~. ~.'

. _ r.. .)d.. "...a.

Design Transitien Temperature (DTT), which is equal to ZTT + 60 F, the pressu e

. e s e el_e v411 %~ c, e - a' e ' s o ".. s',. 'e .". a. e * *. e e s .' e "< e '. e - _' .' %

. - ". e . . .4,~. e* _=

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r -,.

e. .

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. m. -;,. t, .e.

-- cc 0 :. , c- _- r. : . ,,

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,-a_ rce u +e "

v.4 a_ l d-

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-n.

r,,,

+. k..e +v e_.- r r.c- ,. . . . . e cci *u ~v- _ i. ~, , e. . . e n, O r +n .?~".,

p e

_-.f.:..- _ . ~ .e .*.w. _ .e e k. <. n . .

vill increase frc: 10 to 20 percent yield strength.

.. p +. 3. ,. + e .y - g. . ~. 3. , , e c '. ..'.'.' , a . ax .i . "u. s -ae-.-_a c '. 20 p e ~ e ~~ . . v, .' r. .i d -

.c. . .. e ~ -+.

e . %. .

, e. .v_ .a a

.e.. n _s ,. . . a ,e .n.., .= . -..<

. n..... +. n. . . .n. e .e. e.-. -. a .._n.-=- . . . . . _ . . - ',e._

.-....a..

e - a. .,. _ ,\ ,

> .4..,. . . a , . , , . e v.4 1.' +c

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.~ ~ yv

. ~ .~. s. w ., -

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i;> -v>1: ".

.. .. - m..?- m .. a .v ... e . . A.

ne

+

- _s . s_e

.e> ...-

. . , . . ._w..

..s. . . .. n_ .e .c., *4

. ~..

e_ o. c. 5. ,_ C .... .. . -1.c.4. v. c. .._.y e - , e.. :- .-. _ .o ..

7

...t

--.a....m.

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r *.. m C (. %%D

i. ~ .t

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• * - . ~

....*4_ ... .

,v

,., .-,.]..,

l t

/

,g that control pressure and te=perature during heatup and cecidown.(3) This pro-7_) cedure vill insure that the stress levels do not exceed those specified in a

, through e above.

1' h.l.h.h Quality Manufacture i

The quality centrol pr.cgram for the reacter ecolant syster is as cutlined in L.5 This progre= will be crganized, imp'erented and monitered as descrited in Sl Appendix 13,andAmend:entsNo.6and8.

h.l.5 CODES AND CLASSIFICATIONS All pressure-containing components of the reactor coolant syste= are designed, fab-27 ricated, inspected, and tested to applicable codes listed in Table 4-11, as supple-cented by the Code Cases listed in Table 4-lla.

h.2 SYSTD' DESCRIPTION AND GPERATION L.2.1 1E';E ,AL DESCRIPTION f The reactor coolant syste censists of the reactor vessel, two vertical once-through steam generators, four shaft-sealed coolant circulating pu=ps, an elec-

) -

trically heated pressurizer, and interconnecting piping. The system is arranged l as two heat transport loops, each with two circulating pumps and one steam gen- d g erator. Reactor syster. design data are listed in Tables L-3, 4-h, L-5, L-7 and h-8, 'and a 'sys te= scheratic diagra: is shown in Figure L-1. Elevation and plan views of the arrangement of the major cc=ponents are shown in g Figures h-2 and h-3 4.2.2- MAJOR COMPONENTS l h.2.2.1 Reactor Vessel The reacter vessel consists of a cylindrical shell, a cylindrical support skirt,

a spherically dished botto: head, and a ring flange to which a re=cvable reacter

. closure head is bolted. .The reactor c1csure head is a spherically. dished head velded to a ring flange.

The vessei has six major no::les for reactor coolant flew, 69 control rod drive nc :les =cunted on the reactor closure head, and two ccre flooding nc :les -

all located above the cere. The reacter. vessel vill be vented thrcugh the cen-trol rod drives. The vessel closure seal is forned by two concentric C-ring:

1

.vith provisiens between them for leakage collecticn. The reactor vessel, nc::le' design, and seals incorporate the extensive design and fabrication ex-perience accumulated by 3&~a'. Forty-six in-core instrumentation no::les are located on.the 1cwer head.

'The reacter closure head and the reacter vessel flange are joined by sixty 6-1/2 in. diameter studs. Two metallic 0-rings seal the reacter vessel when the reac-ter closure head.is-belted ~in place. Leakoff and test taps are provided in the annulus between the two C-rings to dispese of leakage and to hydretest the.ves-sel c1csure seal' arter refueling.

(j T*.e ve:sel is insulated with metallic reflective-type insulatien. ~nsulatien ranels are previded for the reacter elesure head.

0000G' L _ :'

.Amerd ant No. Ic'

~ _ ., .

4/75 . - .

g The reactor vessel internals are designed to direct the coolant flow, support y the reactor core, and guide the centrol rods in the withdrawn position. The reactor vessel contains the core support assembly, upper plenus assembly, fuel assemblies, centrol rod asse blies, surveillance specirens, and inccre inctru-mentaticn.

The reactor vessel shell caterial is protected against fast neutro. flu). and gn- a heating effects by a series of water annuli and the therral shield located between the core and vessel vall. This protection is further descrited in 3.2.h.1.2, k.l.h, and h.3.1.

Stcp blocks velded to tne reacter vessel inside vall limit reacter internals and core vertical drop to 1/2 in. cr less and prevent rotatien about the verti-cal axis in the unlikely event of a major internals cc pcnent failure.

Surveillance specimens made frc reactor steel are located between the reacter j vessel vall and the therral shield for Unit 1 and Unit 2. These specimens will be examined at selected intervals to evaluate reactor vessel caterial NI7PT changes as described in k.L.3 The reactor vessel general arrangement is shown in Figure h-5, and the general arrangement of the reacter vessel and internals is shown in Figures 3 LS and 3-49 l Reacter vessel design data are listed in Table L-3 k.2.2.2 Fressurizer The general arrange:ent of the reactor coolant systen pressurizer is shcun in

} l Figure L-6, and the design characteristics are tatulated in Table L L. The electrically heated pressurizer establishes and maintains the reacter coolant pressure within prescribed lhrits and provides a surge chamber and a water re-

- serve to acce==odate reactor coolant volume changes during operation.

The pressurizer is a vertical cylindrical vessel connected to the reacter cutlet piping by the surge piping. The pressurizer vessel is protected frc ther:21 effects by a thermal sleeve on the surge line and by a distributien baffle located above the surge pipe entrance to the vessel.

Two AEMI Code relief valves are counted on the tcp cf the pressuriner and func-tion to relieve any syste: cverpressure. Each valve has one half the required relieving capacity. The capacity of these valves is discussed in L.3.L. An additicnal pilot-cperated relief valve i provided to limit the lifting fre-quency cf the code relief valtes. The rel:ef valves discharge to a pressuriner quench tank located within the reacter building. The cuench tanP ".a a s:cred water supply to ecndense the stea=. A rupture disc prctects the tani against overpressure should a pressurizer valve fail to reseat.

The pressurizer centains replaceable electric heaters in its 1cwer section and a water spray nozzle in its upper section to ruintain the steam and vater at the saturaticn temperature correspending to the desired reactor coolant syster pres-sure. During outsurges, as the pressure in the reactcr decrcases, 2:re cf the water in the pressurizer flashes to steam to maintain pressure. Electric heaters are actuated to restcre the ncr:a1 cperating pressure. During insurges, as  ;

(^'.  !

C

. ~ r A~Mni? Int No. b L-6 . \

2 lo.o

- 70 :

l

pressure in the reactor syster increases, steam is ecndensed by a water spray O frc a reactor inlet line, thus reducing pressure.

are centrolled by the pressuriner pressure centro 11er.

Spray flow ana heaters Instruzentaticn for the pressurizer is discussed in 7.3.2.

h.2.2.3 Stet: Generator The general arrangement cf the stea a eneraters is shcvr in Figure L-7, and de-l sign data are tabulated in Table L-5

The stear generator is =. vertical, straight-tube-and-shell heat exchanger and

. produces superheated steam at ecnstant pressure over the power range. Eeactor

' coolant ficvs dernvard throuch the tubes, and steam is cenerated cn the shell side. The high pressure parts of the unit are the hemispherical heads, the tube-sheets, and the straight Incenel(*) tubes between the trbesheets. Tube supports held the tubes in a unifer= pattern alcng their length.

The shell, the cutside of the tubes, and the tubesheets fer: the bcuniaries of the stear-producing section of the vessel. Within the shell, the tute bundle is surrounded by a baffle, which is divided into two sections. The upper par; cf i

the annulus between the shell and baffle is the superheater outlet, and t' lever part is the feedvater inlet-heating cone. Vents, draina, instrumentatic:

no::les, and inspection openings are prcvided en the shell side of the unit.

The reactor coolant side has ranvays en both heads, and a drain no :le fcr the botto head. Venting of the reactor coolant side cf the unit is acec plished by a vent connection en the reactor coolant inlet pipe to each unit. The unit

() is supported by a skirt attached to the buttc: head.

.Eeactor coolant water enters the stear generater at the upper plenum, .lcvs devr the Inconel tubes while transferring heat to the seccndary shell-side fluid, and exits through the lever plenu=. Figure L-8 shcws the f1cv paths and steam gen-erater heating regions.

Four heat transfer regicns exist in the stest generater as feedvater is ccnverted te superheated steam. Starting with the feedvater inlet these are:

a. Feedvater Heating Feedvater is heated to saturaticn temperature by direct ccntact heat exchange. The feedvater enterin6- the unit is sprayed inte a feed-heating annulus (downecter) fcnned by the shell and the tarfle arcund the tube bundle. The steam that heats the feedvater to saturaticn is drawn into the devnecter by cendensing acticn cf the relatively ccid feedvater.

(* )Inconel is a trade nace cf an allcy snufactured t 3 the Ir ernatienal ickel Cc pany. It also has substantial cc ron usage as a generic descripticn ~ a A '

' i-Te-Cr alley ccnferring te ASTM Specification SE-163 It is in the atter centex ,n at reference is nn:e nere.

i -

4-

$79.hMT.1 50. c g 2 /~c

1 l

l

. l

+n v8.a

". .e. a. s a.. t. a+w e A =. . a+,, e , 4..e. +vwue Aw C.. .e.. n C- a. .,.v.re .n e a e+ a+ 4. n v. . a. ., A. 1 .. .n e

) - . . ..

I

+. k.. o. 3..g+. i n. h o_ n_m A. 4 e.. +. k. . e .c.u- a l e a+. e k 4

& c 1 4 .e.3. s e c 4r

+m.c. . P

. b. . .i e. ,--ra..'

.e c o. .e +>e i

he=d to overecce pressure drop in the circuit fer.ei t. the dcun- ,

cc_er, the boiling sections, and the types stean ficw tc the feed-vater heating regicn. "ith lov (less than 1 ft/sec) saturated water velccities entering the generating section, the seccndary side pres-in t.ne bolline sertlen are ne m cicle. ..ne can. crits" c.,

sure crc-s r - -

the pressure drop is due to the static head cf the nixture. Ccnse-cuently, the devnconer level of water balances the sean density of the two-phase boiling sixture in the nucleate boiling regicn.

b. Nucleate Eoiling The saturated water enters the tube bundle, and the stea -water six-C.c +w ...e ..o.., .w.. .s. o. .e ,,c..e. .

- - . . l c .. .e.

,,..,.A of a c.. .

C.. . +he C,1.e4Ae ...w w... . m... ,,..e.. ..

'.k..e .-C a * *.

• r C C ol o. ..*. .. #' c". . ~~. ne Va"y .- C o . .*s e n" +. c #. . 4. 7.*. -

.".~.a.....,,"u.e.-

. . .ceeee. ,1_ . . .e. . ,w. . . 4 c

.e. ,.y.

up.+. 4 1

.. . M...~, ..e.,

A c. y. c. .+ . m.

g-.e.o.

v. .,. 4 .1 4. o3 e , 4. e .a a n y..a.- A , av.a. +wh.e-

. . ~ . r4. 3. %v c . ;..12 ..e g.e.4 - s w-yo. .- v. . a.c.. .. ; r. . e w~.m~ n v.e .c .

4 .. e qu_,. .. , 4 ..a.. g.,, .k. .' c 'n *o.

• a.".s .4 'e i c .1 2'.c. ..- la.a*.e k.4'4.e-

"a

. . . . ,s _

. ' - b d' .i .e~

j cecurs is a function of pressure, iu.:at flux, and cass velocity.

C. fl..,O colling Era saturated steen is -roduced r ir t k a '4'- teiling regien at the un.r.er end cf the tube bundle.

A .c .. , .~ . v~ e a+. ,. A .c + e a Saturated stean is raised to final terperature in the superheater e -z3.c. .

c<. >..ea.:.3e..- s ,~... ,a . e e.m ., c ,~., ,, , ,. ,.,..e...e , v e ,. .s , , s , m. a A_ .

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f. .r- e'2 .c .a.' .o e

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y... e..,

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. v... .....

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. '* 6"O

s. . c._ s .

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f'cv characteristics en both the reactor ecolant and seccnda y sides. "he hct res.cter coolant fluid is eccled uniform 1y as it ficvs downward. The seccndarv .

sic e rass flev is lov, and the majority of the pressure drop is due to the static effect of the nixture. The boiling in the stear generator is screwhat sirilar S,1,.l u e, n ex e.. w . .w.,e 3 4C... ._,.A . s. . .e. ., . e. . e cee,..,,,,.-....

.N, . -

+wv..a. . . . is .. w ... --

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r A r.ne. .. c . . ..e.. . a. anu e . C. c. .

t C.

y..,' ) ;r ,.

Lv

7J A plot of reactor coolant and steam temperatures versus reactor pcwer -is shewn in Figure 7-5. As shown, both steam pressure and average reactor coolant ten-perature are held constant over the load range from 15 to ICO percent rated q power. Constant steam pressure is obtained by a variable two phase boiling U .'ength (see Figure 4-8) and by the regulation of feed flew to obtain proper steam generator secondary cass inventory. In addition to averace reacter cool-ant temperature, reactor coolant flow is also held constant. The difference between reactor coolant inlet and outlet te=peratures increases proportionately j as Icad is increased. Saturation pressure and temperature are constant, re-sulting in a variable superheater outlet *ceperature.

j Figure 4-10, a plot of tc=perature versus :ube length, s5cus the tc=perature

dif ferences between shell and tube throug

t the stear generator at rated load.

The excellent heat transfer coef ficients pe rmit the use of a secondary operating pressure and temperature sufficiently close to the reactor coolant average ter-perature so that a straight-tube design can be used.

The shell temperature is controlled by the use of direct contact stean that i heats the feedwater to saturation, and the shell is bathed with caturated water fron feedwater inlet to the icwer tubesheet.

In the superheater section, the tube wall temperature approaches the reactor coolant fluid temperature since the steam film heat transfer coefficient is considerably lower than the reactor coolant heat transfer coefficient. By baffle arrangement in the superheater section, the shell section is bathed with super-heated stea= above the steam outlet nozzle, further reducing temperature dif f erentiali between tubes and shell.

t s

The steam generator design and stress analysis will be perforned in accordance

) with the requirements of the ASME III as described in 4. 3.1.1.

},

I 4.2.2.4 Reactor Coolant Purps l~ The general arrangenent of a reactor coolant pump is shown in Figure 4-11, and j 8l the pu=p design data are tabulated in Table 4-7. The reactor coolant pumps are vertical, single-speed, shaf t-scaled units having bottoa suction and horizontal i discharge. Each pu=p has a separate, single-speed, top-rounted rotor, which is connected to the pump by a shaf t coupling.

t Shaf t sealing is accomplished in the upper part of the pump housing using a throttle bushing, a seal chanber, a rechanical seal, and a drain chamber in series. In crder to provide a reliable source of cooling water for the reactor coolant purp shaf t 26 seal, seal injection uater is supplied from the rakeup and purification (MU?) system through ASME Eoil'er and Pressure Vessel Code,Section III Class 3, Seisnic Category I piping. Seal water is injected ahead of the throttle bushing at a pressure approxi=

nately 50 psi above reactor systen pressure. Part of the seal flev passes into the pu=p volute through the radial pump bearing. The recainder flcus out along the thrott j bushing, where its pressure is reduced, to the seal charber and is returned to the ,

j seal water supply syster. The outboard mechanical seal normally operates at a pressuq

, and temperature of approxicately 50 psig and 125 F.11cwever, it is designed for full 26l reactor coolant systen pressure. The outboard drain chanber wculd further prevent leakage to the reacter building if deterioration of the rechanical seal perfornance should occur. -

n 4-9 Arcndnent *:o . 26 L!I j

A water-lubricated, self-aligning, radial bearing is located in the pump housing.

An oil-lubricated, radial bearing and a Kingsbury type, double-acting, oil-lubricated thrust bearing are located in the pu=p notor. The thrust bearing is designed so that reverse rotation of the shaf t will not 1 cad to pt=0 or rotor

.l datage. Lube oil cooling is acco plished by cooling coils in the :: tor oil reservoir. Oil pressure required for bearing lubrication is maintaired by internal purping provisions in the notor, or by an external syste if required for " hydraulic-jacking" of the bearing surfaces for startup.

1 1

An antirotation devi:e will be furnished with each purp : tor to pr hibit re-verse rotation of the pu p.

Factory thrust, vibration, and seal performance tests will be cade in a closed loop on the first pump at rated speed with the purp end at rated terperature and pressure. Sufficient testing will be done on subsequent units to substantiate that they conform to the initial test pt p characteristics.

' An analysis has been perfor ed to deter .ine the consequences of the loss of Component Ccoler Water (CG?) to the reactor coolant purps.

1. A loss of CC: to the seal jacket cooler will not preclude cooling of the pu=p seals as the seal injection water will continue to act as a heat sink in passing through the seals.

2. A loss of CCJ to the seal return cooler will not affect opera' . :n of the

" reactor coolant pumps. The heat picked up by the seal injection water vhen passing through the pump seals even without seal water cooling which would be afforded by the seal jacket cooler or the seal return O.

cooler is returned to the nakeup tank. The nakeup tank acts as a heat 4

sink for the one gallon per minute hot return seal water. Eere the te: perature of the nakeup tank is unaffected by the hotter retrurn seal 26 water. This is due to the large volu e of water in the tan and also due te the cixing in the tank of other cooler influent streams.

~

3. Upon loss of CCJ to the notor lube oil cooler, the noter 1:5e oil te -

perature will rise and the reactor coolant pu .p notor cust be de-energized wit 11n 19.4 =inutes of less of CCf to insure a da age-free coastdown.

All of the f( regoing situations are alarred in the control roo: signaling the operator to take corrective action as required. In the case of loss of CCi to the

reactor coolant purp notor lube oil cooler, the operator is warned by a high oil terperature alarm in suf ficient time to allnw de-cnargizing of the pu p. High teep-

, erature alar s are also located on the CC;? return frc the seal jacier cooler, the notor lube oil cooler and the seal return cooler. Also the seal return water exiti6g a

the pump seals is alarmed for high pressure and low pressure.

4 j In the unlikely event that component cooling water should be lost to all four l reactsr calant purp notor lube oil coolers, the operator would trip the reactor upon noting the high oil temperature alarm. Prior to 19 ninutes af ter noting the alar:

the operator would de-energize all four reactor coolant pu ps. The pu ms vill coast

' down to a standstill in about three Linutes. A saf e shutdcun to cc:'fitiens of pressure and temperature for decay seat rcroval initiatien will bc raintained by use of the auxiliary or rain feedwater syste .s. Both of these feeduster sys ters have the f

ability to induce natural circulation of the reacto: ccolcm sys:c . ::o fuel de ye will result f rc-- this natural circulation made of cc c'..m a c _ m: in the l relysis fer less cf electric power accident in Sectica 1 .

4 4-10 yn . 'A A

• e M c-t ':c. 26 Ll%

v i

i i a , ,

b l

1 Reactor Coolant Piping

] 4.2.2.5 The general arrangement of the reactor coolant systc= piping is shcun in Figures 4-2 and 4-3. Piping design data are presented in Table 4-E. In addition to the pressuricer surge piping connection, the piping is equipped with welded ccnnec-tions for pressure taps, terperature elements, vents, drains, denay heat rer.c val ,,

and energency core ecoling high-pressure injection water. Ther a. sleeves are provided in the pressuricer surge piping, the energency high pressure injection.

a and the core flooding connections.

a 4.2.3 FRESSURI-RILIEVI!;G DEVICES 7

The reactor coolant systen is protected against overpressure by centrol and pro-tective circuits such as the high-pressure trip and code relief valves located eq the top head of the pressuricer. The relief valves discharge into the pressuricef

~

j quench tank which condenses and collects the ef fluent. The schc atic arrangenent cf the relief devices is shown in Figure 4-1. Since all sources of heat in the systen, i.e., cere, pressuricer heaters, and reactor ccolant pumps, are inter- al) connected by the reacter coolant piping with no intervening isolation valves, relief protectica can conveniently be located en the pressuricer.

4.2.4 E' TIRO 0E?;TAL FROTECTIO:;

The reactor coolant systen is surrounded by concrete shield walls as described j in Section 5.

eleva-J Lateral bracing will be provided near the steam generator upper tubesheet

-* d (3

U tion to resist lateral loads, including those resulting frca seistic forces , pip

I rup ture ,

therral expansion, etc. Additional bracing is provided at a lower eleva

o. .

tion to restrict whipping of the 36-in. ID, vertical pipe leg.

1 I

k 1

k E

4 7

(i

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1-103 Ond'ent '! O . 16 r-,

m ,w

T 1

4 h.2.5 MATEFIALS OF CONSTRUCTION h.2.5.1 General l Each of the caterials used in the reactor =clant systen has been riected for the expected enviren= ental and service cc=E:icns. ' he rajor et _ "rt mate-lrialsalclistedinTableh-12.

All reactor coolant systet materials exposd to the coolant are =- csion-

~

resistant materials consisting of 30h or M SS, veld deposit 30-55 cladding, a4 Incenel (Ei-Cr-Fe)., and l'T b PH (H1100). rese caterials were c=ce for s

specific purposes at varicus locations vitii: the systen because ::' their su-

[! pericr cc=patioility with the reactor cooW.

l Periodic analyses of the coolant chemical == position vill be pe '=ed to ten-itor the adherence of the system to the ree=or coolant water q.2.' y listed in

~1 l Table h-2. Maintenance of the water quali y to minitice corrosi= s performed by the chemical additicn and sa pling syster which are descrileiin detail in 9.2 and 9.10.

The feedvater quality entering the stea: grerator vill be held c__i the lit-M l its listed in Table h-3 to prevent deposits and corrosien inside:te steas gen-7 erators. This required feedvater quality izs been successfully esed in cacpar-

! able ence-through, nonnuclear stea= geners =s.

All external insulation of reacter ecclant systen cc penents vil' be ec:patible

% vith the ec=nonent raterials. The reactor wssel is insulated .2:etallic i h reflective insulaticn on the cylindrical M extericr. The cl: c e flanges and the top and botto= heads vill be 'insubrd. vith lev-halide-- insulat-

-^M ing raterial. All other external corresian esistant surfaces i :te reactor M

coolant system vill be insulated with lev-in.Ilde insulating cate 's' as required.

h.2.5 2 Pressure Vessels l

Tae reactor vessel plate =aterial opposite te core is purchased:e =btain a Charpy V-notch test result of. 30 ft-lb or g=ater at a ccrrespoA Kil-Iuctility Transition Temperature (NITIT) of s nc inal value of 107, and the

=aterial aill be tested to determine this t==al NITIT value. In =" tion, this J

plat ~e vill be 100 percent volumetrically 1 rected by ultrasonic t=st using

-both nornal and shear wave.

I t._ The reactor vessel caterial is heat-treated specifically to obtt': g o:1 netch-l ductility which will insure a lov TDTI and .ierecy give assurance :ist the fin-ished vessel can be initially hydrostatice-_; :estec anc cperatei s: rec te -

perature without restrictions. The stress ~'-its established fc ne reactor

,a, vessel are dependent upon the temperature at which the stresses re applied.

I As 'a result of fr.st neutron absorption in te region of the core, tie caterial i ductility vill change. The effect is an i= ease in the NDTT. 'ie p edicted I

end-of-life NDTT value of the' reactor vessi cpposite the core, irei en an

' initial value of 10 F, is 260 F. The predi=ed neutrcn exposure ri D'~ shift L} are discussed in L.l.k. ,

1

..-4

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I

h.2 7 LEAK DETECTICS D

8 To minizine leakage frc: the reacter ecolant systen all ec pcnents are inter-connected by an all-velded piping syster. Scre of the ec ponents tsve necess cpenings of a flanged-gasketed design. The largest cf tiiese is the reacter closure head, which has a double metal 0-ring seal wit.. provisions for dis-posing of leakage between the 0-rings.

'a'ith regard to the reacter vessel, the probability of a 2eal occurring is con-sidered to be re=cte cn the basis of reactor vessel design, fabrication, test, inspecticn, and operation at temperatures above the material NDIT as described in L.3.1. Reactor closure head lealage vill be zero frc: the annulus between the =etallic O-ring seals during vessel steady-state and virtually all transie operating conditions. Only in the event of a rapid transient cperation, such as u trgency ecoldern, would there be sc=e leakage past the innernest 0-ring seal. A stress analysis en a similar vessel design indicates this leak rate vould be apprcximately 10 cc/ in through the ceal =cnitoring taps to a drain, and nc leakage vill occur past the cuter C-ring seal. The exact nature of this transient condition and the resulting small leak rate vill be determined by a detailed stress analysis.

In the unlikely event that significant leakage should occur frc: the systen into the reactor building during reactor operation, the leakage vill be de-tected by one or : ore of the following =ethods:

a. Instrumentation in the control rect vill indicate the additicn rata cf nakeup water required to maintain normal vater level in the pres-suriner and in the takeup tank. Deviation frcs ncr:al takeup and letdown to the reactor coolant syste vill provide an indication cf the magnitude of the leak.

t. Ccntrol roc: instruzentation vill indicate an increase in rea .cr building activity.

c. Radvaste instrumentation vill indicate the existence cf a large a: cunt of vater flow frc: the reacter building surp.

L.3 SYSTEM DEslGN EVALUATIc; c.- y.

i

.1

• . 3 - .s: rn, c. i c r.o The reactcr eccla;t systen is designed, fabricated, and erected in acccrdance with preven and reccgnized design codes and quality standards applicable fcr the specific con; cnent functicn cr classificatien. Inese cceponents are ce-signed fcr a pre ssure of 2,500 psis at a nominal temperature of 650 F. The corresponding ncminal operating pressure of 2,185 psig allows an adequate margin for normal lead changes and operating transients. The reactor system compcnents 2? are designed to =cet the codes listed in Table 4-11, as supplerented by the Code Cases listed in Table 4-lla.

! side fret the safety facters introduced by code requiretents - _ quality con-trol programs, as described in the following paragraphs, the reacter ecolant

\s 1 systen functional safety facters are discussed in Secticns 3 and 14 ec, ourts.- . .

L ',=" ":E n :c e r.: .';c . 2c 4/75

i , y w ,rn

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--a . . . _ .

closure head, which has a double retal C-ring seal with : revisions fcr dis-

r. esin5 of leakare between the 0-rines. - -

'4 *. +. . b. . e e g.*A- *n 6v .wk. . n. w e s +...n. y e e c o. l , +.k.. e. ,.C % k'.i4.-

. . . . , c ,. e. _' n_ c '_..~:.... . . 4. _- .. - . . _

sidered to be re=cte on the basis of reacter vessel design, f atrication, est, inspecticn, and operation at terperatu es above the raterial HD T as described in L.3.1. Reactor closure head leakage vill be cerc frc the annulus between the retallic C-ring seals durino vessel steadv-state and virtually all transient . .

~, e-a+. i..e c c .^. .4.*. 4. c. .s . O.n1."7 .i n.. +. k..e e". e a*e c'a . a"_d . a . . .c .' a . c , a. c- ". . c , .n ". - k..

4 . . , -. . .

as an ener eency eccidcun, vculd there be scre leakar_e tast the innernest C-ring .

seal. A stress analysis en a similar vessel design indicates this leak rate

..~.~,,a

.. a b e a-,. , x 4 - o.+. e.' v. .'.O c c l 4 . ' k. - c"u e-*.

• k.. a. eeu'c. -. . #. '.~r .' . . e a. ~, .~ c- d..-='...,

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^

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- 1. 4 .e .e.. c .' _'

'a=> .='a. ' . " _ ' _ '%. ud a- *. a. . _. 4. . . M. . k" a detailed stress analysis.

1. T

.. . .g. e up..,.4.F.al"s ev e ' 'uh c'. .ed-.4'ica

.e. a ' 'eak.a3-a

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i d c - ~- ~. . . '- _. 'e 'n e .e .". .e .

• e_.

into the reacter buildine durine reactor eneration, the leakage vill be de-tected by one or more of the following methods:

a. Instrumentation in the control rec vill indicate the addition rate cf takeup water required to raintain nonal water level in the pres-O sur4..e- . an'..

4-a 'e k e .= >. .e " y *m= a- >.. .

letdevn to the reacter ecclant syster vill prcvide ar indication of n.~a".'c* m n .'.. .~. ._.='

4

. _ . c-t.a." , c...' --

the magnitude of the leak.

b. Centrol roc instr =entation vill indicate an increase in reacter building activit~. J

c. Radvaste instrumentation vill indicate the existence of a large arount of water flev frc the reactor building s =p.

i ev. e 1vv4n.r.~e w v T c- 3 mm l,n.1U

.uu o .

4.. s .

1 e,:L1.y w n. f,AC A Cpq

.6 Tne reacter ecolant systen is designed, fabricated, : nd erected in acccriance k'

...4 'e'n p". c". e n c- ".d ."e C C en#. a. d u# e.. i e'*.". " C # a. e ="..#- ce" ' 'm# ' ,". .e .'. S . #. '_ _# - "ay,

. - -A. .# ^ 5.- a. .#...-

+. k.. a. n cmeo.imr4 C C C ,~* *. A_ . . . "A'. .#Un* tic.". cr CI*ee##.#C

. . m m" *~4 - .. . ".~"a..e.=. .----.e..+,..e.

. a-a. e=.-

signed for a pressure of 2,500 psig at a nceinal te peratu e of 650 F. ne --

' corres-cnding r ncs.inal operatins pressure of 2,165 sig r a11cvs an adequate car-gin for nr: al lead changes and cperating transients. The reacter syste: cc -

I'enentsaredesignedtoreetthecodeslistedinTableL-ll.

p

.aa e .r. c. w .. .+ ne. ,- re + , . r, n. s. o.na < a.-a a u_e a v- w: c~a , ~~+27 m. .. . . . . , a . ~a c ,.,<.-

c. e . .u . . . . . . . . . . ..; c-..-

trol trcerans, as dese:ibed in the fellcvine r.ararraehs, tr.e reacter ccciant e . . e . e.

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00016  : ..

g ,; . . . . .x_

. . . . e. .. . . .. . , ,

E r*.

k l

l 4.3.1.1 pressure vessel (Including Stes= Generator) Safety O The safety of the nuclear reactor vessel and other reactor coclant systes pressure vessels, including the reactor coolant pump casing, is depenient upcn fcur =ajcr facters: (a) design and stress analysis, (b) quality control, (c) proper operation, and (d) additional safety factors. The special care and detail used in i=plementing F these factors in pressure vessel =anufacture are briefly described as fc11cvs:

- k . 3.1. l ~.1 Design and Stress Analysis p These. pressure vessels are designed to the require ents of the ASME III code.

This code is a result of ten years of effort by representatives frc= industry

i. and govern =ent who are skilled in the design and fabrication of pressure ves-f sels. It is a ec prehensive code based on the most applicab3e stress theory.

l It requires a stress analysis of the entire vessel under both steady-state and transient operations. The result is a ec plete evaluatien of both pri=ary and I

secondary stresses, and the fatigue life of the entire vessel. This is a con-trast with previous codes which basically established a vessel thickness dur-ing steady-state operatiens only. In establishing the fatigue life of these J. l pressure vessels, using the design cycles from Table h-9, the fatigue evalua-

.) tion curves.of ASME III are e= ployed.

Since ASME III requires a ec=plete stress analysis, the designer must have at his disposal the necessary analytical tools to accceplish this. These tools

'~ are the solutions to the basic =athematical theory of elasticity equations.

7 In recent years the capability and use of ec=puters have played a =ajor part p, in refining these analytical solutiens. The ELV Cc=pany has confir=ed the v theory of p3- as and shells by reasuring strains and rotatiens en the large i

flanges of a aal pressure vessels and finding them to be in agreement with those predictud by the theory. ELW has also conducted labcratory deflection '

studies of_ thick shell and ring combinatiens to define the accuracy of the theory, and is using cc=puter programs developed on the basis of this tert data.

The analytical procedure considers 'all process operation conditions s A detailed design and analysis of every part of each Section III vessel is prepared as follevs: 3 i

a.

The vessel size and configuration are set to neet the process require- l

=ents, the thickness requirements due to pressure and cther structural '

dead and live leads, and the special fillet contour and transition taper require =ents at no: les, etc, required by ASME III.

b. The vessel pressure and temperature design transients given in Table h-9 are .e ployed in the determination of the pressure loading and ltemperaturegradient.ndtheirvariationswithtimethroughoutthe vessel. The resulting co=binatiens of pressure loading and thernal stresses 1 are calculated. Cc=puter programs are used in this develop- j

. =ent.

c. The stresses through the vessel are evaluated using as criteria the -i allowable stresses per ASME III._ This' code gives safe stress level'

' tait's;for all the types of applied stress. These are setbran'e stress  !

(to insure adequate tensile strength of the vessel), re=trane plus i b,,

L -l' .. k(. -((

Amendnent no..s  :

2/9 '70

i pri=ary bending stress (to insure a distcrtien-free vessel), secon-

~ ,f - dary stress (to insure a vessel that vill nct preg essively defc=

] under cyclic leadirg), and peak stresses (tc insure a vessel cf j naximu= fatigue life).

A design report is prepared and submitted to the jurisdictiencl authorit' s and regulatory a; cachs, i.e. , state , insurance, etc. "his repc-t defir in sufficient detail the design basis, loading ecnditions, etc, and vili

I surrarize the ccnclusiens to per=it independent checking by interested parties.

In evaluating the integrity of the reactor ecolant syster, it is important to consider that the size cf the defect in a syste: ec:;onent that can conceivably a contribute to the rupture of a vessel depends en the sice and crientatien of the defect, the =a6nitude of ine stress field, and terrerature. Tnese rajor parameters have been correlated by Fellini and Fucak(2) who have prepared a

" Fracture Analysis Diagra " which is the basis of vessel operaticn frc: cold startup and shutdcun to full pressure and terperature Operaticn.

Tne diagrs= predicts that , for a given level of stress, larger flav sizes vill

,;, be required for fracture initiation above the NDTI t'emperature. For examole, at stresses in the order of 3/h yield strength, a flav in the order of 8 ' o 10 t

in. =ay be sufficient to initiate fracture at temperatures belev the NDTI te:-

perature. Ecvever, at KDTT +30 F, a flav of 1-1/2 times this size may be re-quired for initiatic:5 of fracture, while at a terperature of IDTZ +60 F, brittle fracture is not possible under elastic stresses because brittle fracture prepa-

% gation does nct take place at this te=perature. Fractures abcve this tempera-

[

~

ture are of the predecinantly ductile type and are dependent upcn the = ember

.q net section area and section =cdulus as they establish the applied stress.

BEN has released a TcF. cal Report covering the reactor vessel's capability to withstand the ther=al shock of'the ECCS vater icilcving a IDCA vithout loss of structural integrity.

1 Tnis report concludes that the reactor vessel vill maintain its integrity de-spite the potential limited crack prcpagation due to the ther..al shock. Tnus ,

based upon this investigation, a syste to ritigate the cenceqcences of a

, vessel. failure due to themal shock folleving a LOCA is not justified.

Due to the proximate population. density, however, previsions vil be added.

to the building and syste=s designs to perrit the pessible future addition of a syster to assure continued core cooling if . info =atien beco es available in the future to. demonstrate that such a system is required.

1 Reactor Vecsel Closure Studs

}

The reactor closure head is attached to the reactor vessel with sixty, 6-1/2

i. . .

in, diameter studs. .Tne stud =aterial is A-540, Grade 323 (Asz III, case

.[ 1335) which has a =ini=u. yield ~ strength of 130,000 ;ci. Tne studs, when tightened for operating ecnditions, vill have a tensile stress of cpproxi-nately 30,000 ps1. Thus, at operating conditiens (2,155 psig):

9 h j 00W8 I l. 'A end ent No. 5 j 11/3/@

a

a. 10 adjacent studs can fail before a leak occurs.

h, 25 adjacent studs can fail before the rensining studs reach yielf. strength,

c. 26 adjacent studs can fail before the remaining studs reach the ultimate tensi e strength,

d. L3 synnetrically located studs can fail before the recaining studs reach yield strength.

The stress analysis of the studs vill include a fatigue evaluatien. It is not expected that fatigue evaluation vill yield a significantly high usage facter for the LC-year design life.

A d

pm I

N d O' L-15a i AEOL52*El IC- E

. ~-.~

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l l

i 1

1 q

3, Stea Generatcr

).

The basic structural precise of the steam generator is that the tubesheets themselves are cesigned to take the full design pres:ure cn either side ;f the tubesheet with zero pressure on the other side. That is, the tubes are nct counted upon for any structural aid er sur.rert.

Differential thermal exrancion- .

induced stresses in either the tubes or shell have been evaluated using the l cost severe design transients from Table L c.

The steam line failure analyzed in 1h.1.2.9 clcsely simulates this design pre -

ire in a transient canner. Secondary temperature variaticns during the acci-sien+. are essentially transient skin effects with the controlling terperature for the tubesheets and tubes being that of the reacter coclant. Thernal stresses for this case vill be belev AS:G alicvable values. Scre tube deferraticn cay occur but vill be restrained by the tube suppcrts.

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.  %..mL,. .

generator by 10 to 20 F depending en lead. The effect is to put the tubes in a slight ec pression of 3,000 psi at the 20 F sxinu te.nperature difference.

During startup and shutdown operations the tubes are better than the shell of the stea: generater by LO F. This places the tubes in a ec pressive stress cf 6,000 psi. Thus, the stress levels develcped during normal startup cr shutdc-T. cpera-tiens cause no adverse effect on the tubes since these stresses are well telcv the allcvable stress of 23,000 psi fcr S5-163 raterial. Euckling cf the tutes does not occur since they are supported laterally at LC-in. in,ervals along their len-th. a Tc demonstrate the structu-al adequacy cf the stear generater at this condition, a laboratory unit was constructed of the same tube size, length, and sterial as the steen generator, but of seven tubes in nutter.

It was structurally tested with a thermal difference between shell and tube of 50 F rer 2,000 cycles. This severe thernal cycle test was perferceu with a tube-to-shell temperature differenct twice as great as the taximum expected l during startup and : hutdev: (Transients 1 and 2, Table L-9). Destructive ex-asinatien of the unit after this test indicated no adverse effects frc fatigue, stress, tuckling, or tube-to-tubesheet joint leakage.

L.3.1.1.2 quality Centrol The c.uality ec:.trcl rrce ran for the reactor ecclant syster is as cutlined in c

h.s. This t.rcrran vill be cr anized

. a and 4 'a-anted as described in Ap endix .

l 13 and A .endrents Uc. 6 and 8.

The prinary purpose cf the quality ecntrol procedures and rethods is te 1ccate and determine the size of caterial irregularities so as to a11cv an evaluaticn of defect acceptance, rejection, cr repair. A discussion of flaws and their relationship to =aterial failures is s m.arized in 4.3.1.1.1.

h . ;. . ,. . ., . :.

c . ,. .-e. 4. c..

As , ~vicusly ;entiened in b.l.h, pressure vesse] service life is dependent cn adhe:ence to established operating procedures. '.tessure vessel safety is also

, dependent en prcper vessel cperatien. Therefore, particular attention is given t

v h V O. .# h. .

=

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. . ..  %.... v p la,: t nv

-i -

I to fatigue evaluation of the pressure vessels a.d to the factors that affect e

t fatigue life. The fatigue criteria of ASE III are the bases of designing for fatigue. They are based on fatigue tests of pressu e vessels spcnscred by the AEC and the Pressure W ssel Research C = ittee. The stress linits established for the pressure vessels are dependent upon the tenperature at which the stresses are applied.

As a result of fast neutrcn abscrpticn in the region of the :cre, the reacter vessel caterial ductility will change. The effect is an u.:rease in the :.il-j Ductility Transition Terperature (I;DTT). The deterrination of the predicted

?  : DTT shift is described in h.1.h.l. This ::DTT shift is factored into the unit startup and shutdown procedres so that full operating press = e is not attained until the reactor vessel terperature is above the design transitien terperature I (DTT). Eelow the D~'T the tctal stress in the vessel vall due to both pressure and the asscciated heatup and cooldown transient is restricted to 5,000 - 10,000 psi, which is belev the threshold of concern for safe cperation. These stress levels define an cperating ccclant press =e ter;erature path cr envelcpe for a stated heatup or ecoldevn rate that must te fclleved. Additicnal inic =Etien i

en the determinaticn of the cperating procedres is provided in L.l.L.1, L.l.L.2, and L.l.h.3

,1 h.3.1.1.L Additional Reactor Vessel Sa'fety Factors Additicnal methods and procedures used in reacter vessel design, nct previcusly i

1 =enti ..ed in h.3.1.1 abcve but which are considered conservative and prcvide an g add _;icnal cargin of safety, are as follevs:

. (/

a. Use of a stress cencentratien facter of fcur en assured flavs in I calculating stresses.

l j b. Use of minicu: specified yield strength of the caterial instead of the actual values.

' c. I;eglecting the increase in yield strength resulting rec irradiation

. effects.

> d. The design shift in ?;ITT as given in h.1.'.1 is tased en Exinun pre-dicted flux levels at the reactor vessel 1.: side vall surface, whereas t':e bulk of the reactor vessel raterial vill experience a lesser ex-pesure to radiaticn and cense 1uently a lever chance in ::D'"T cver the life of the vessel.

e. Results frc the method cf neutrcn flut calculations, ne described j in 3.2.2.1.7, have increased the flux calculaticns t. a fe.cter cf tvc 8 in predicting the nyt in the reactor vessel vall. The conservative assu=ptiens , uncertainties , and cczparisons of calculaticnal ecdes f used in determining this facter are discussed in detail in 3.2.2.1.7 j

k.

The fcregoing discussien presents a detailed description cf design, f abrication, and c;erating precedures used to insre ccnfidence in the integrity cf the re-acter ecclant syste: cceponents. Reference 5 and EW experience su;;crt the conclusion that reactor vessel ruptu e is incredible.

e' 00021.

t", (

L.3.1.2 Finirr Eesien, stress analyses. c.uality control and cperating limits for the reactcr ccclant t.in.ine vill provide a level of system intec_rity ec.uivalent tc that achieved for the reacter coolant syster pressure vessels.

Total stresses resulting frc: thernal ex-ansicn r and pressure and techanica] and seismic leadings are all' considered in the design cf the reactor ecclant pipin6-The pressu-icer s r ee line connection and the high press e e iniecticn ecnnecticne o are equipped with therral sleeves to limit stresses frc: thernal shock tc ac-ceptable values. All caterials and fabrication procedures will meet the require-cents of the specified code. All caterial vill be ultrascnically inspected.

All intericr surfaces of the interecnnecting piping are clad with stainicss steel to eliminate corrcsic: problems and to reduce ecclant centaminaticn. pres-sure velds and claddine vill be inspected by ncndestructive techniques includ-ing radicgraphic, dye penetrant, or cagnetic particle examinatiens as appropria._.

e

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., p UAe a-A .- . . . , Vnn i a .wr w-. , c y q- .gr e.,

v.v...

The principal heat renoval systers which are interec.nected with the reacter coolant syster are the stes: and feedvater systers and the decay heat renoval syster. The reacter coolant syster is dependent upcn the steam generators. and the steam, feedvater, and condensate systems fer decay heat rencval frc: n nal cr eratin s ccnditions to a reacter coolant ten .Eture cf ac.crcximatel, 28C F. . . .

All vital active cc penents in these systers are duplicated fcr reliability purpcses.

Flev diagra:s cf the stea and feedve.ter systers are shcun in Figre 10-1. In the event that the condensers are not available to r te ive the stes: generated by decay heat, the water stcred in the feedvater syster may be pumped inte. the steen generaters and the resultant stes: vented to the at csphere. Cne seter-driven, and cne turbine-iriven auxiliary feed pump vill supply water to the steam generators.

The decav. heat re=cval systen is used to rencve decay heat when the therral driving head of the reacter coolant syster is no 1cncer adequate to enerate e steam. This s.ister is ccepletely described in 9.3 - The heat received by .

this syste= is rejected to the ec ponent cooling water systen which also contains sufficient redundancy to insure adequate cperation. A schematic diagram of this syster is presented in Figure 9-8.

L.s.-s .e. e-v. - . _ ., Im.rv.:m

. 4e .my The interrity of the reactor coolant syster is insured by

. ercper caterials selec-tion, fabrication quality control, design, and operation. All cc penents in the reactor ecclant systen are fabricated frc =cterials initially having a lov Nil-Luctility Transition Tenperature (UDTT) to elisinate the possibility of vrceanting-e type failures.

The reacter coclant syste 1s designed in acecrdance with AEM.E pressure vessel and USASI pcVer piping codes as covered in L.1. F.elier valves on the precericer are siced to prevent syster pressure frcs exceeding the design point by =cre than

~ 1C percent.

h h, ,. g .#[y

- f.

L_,v

.A

l i ~

l o As a further assurance of sys'.e: inteErity, all cc penents in the system vill be hydrotested at 3,125 psig before initial operatien. The largest and scst 9

frequently used opening in the reacter coolant syster, i.e., the reacter c1c-sure head, centains provisiens for separate rydrcstatic pressuri:stien between the C-ring type gaskets.

i ) . _o, . L e. n.r SUF.7._577I,77 lhe reacter coolant syste: is protected agai: everpressure by relief valves located cc the top of the pressurizer.

4 The capacity of the pressurizer relief valves is determined frc: considerations of (a) the reactor protection syste=, (b) pipe pressure drcp (static and dy-nacie) between the point of highest pressure in the reactor ecolant syster and the pressurizer, (c) the pressure drop in the safety valve discharge piping, and (d) accident er transient cenditions that may potentially cause cverpres-sure.

preliminary analysis indicates that the hypothetical case cf withdrawal of a regulating centro 3. rod asse bly bank frc: a relatively low power provides the basis fer establi shin 6 pressuriner relief valve capacity. The accident is terzinated by hi; h-pressure reactor trip with resulting turbine trip. This ac-cident ccndition produces a pcver =isratch between the reacter coolant syste:

and stea: syste= lar6er than that caused by a turbine trip without inmediate reacter trip,.or by a partial lead rejecticn frc: full lead.

~

(h The ASMI Section III required relief valve capacity as determined en .the basis

.- of the accident described above is 600,000 lb/h. Two, 300,000 lb/h valves are installed. An additional pilot-operated relief valve, capatie of 100,000 lb/h is provided to limit the lifting frequency of the code relief valv.es.

h.3.5 REDUNDANCY The' reactor coolant syste: centains two stem: generators and four reactor eccl-ant pumps. For added reliability, pcVer to the purps is available frc: tvc separate tuses as shavn in Section 8.<

Separate .ccre flooding no

:les are provided en cpposite sides of the reacter

, Hl vessel to insure core 'reflooding water in the event cf a single ccre flocding -

i 1 no::le failure. Reflooding vater is available frc either the ccre flooding I' I tants er the decay heat pumps. which provide icv-pressure coclant injecticn as

,  ; an engineered safeguards. The high-pressure inf ecticn pipes are cenre ted to j- the reacter ecolant syste en each of the ftur :::lant inlet pipes.

.r

i h;3.6 SAFETY ANALYSIS The ec penents of the ~ reacter ecolant systen are interconnected by an s.ll-velded piping systen. Since the reacter inlet and cutlet no

:les are located

, f

! above the core, there is never any danger cf the reacter ecclant un cvering

the ccre when1 any other syste er pcnent is c~e -a- '-- inspecticn cr repair.

Cc:plete safety analyses are included in Secticn ik.

.O 00023 1

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i

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cf high,,reacter ecolant te.perat re, high presse e,.1cv press e e, and .cv r cv,

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during the service life of the nuclear unit.

Dissi ilar retal velds cd the reacter vessel include cnly the ccre flceding lines, in-core instru=entaticn guide tubes, and centrol red drive heurings.

Dissimilar retal velds in the piping include enly attach ents and the reacter ecolant pump inlets and cutlets.

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valve header, and the spray line cennections.

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Represen;ative icngitudinal and circumferential velds en the piping, steam gen-erator, pressuricer, and pump casing vill be inspectable as described atcve.

Representative velds on the reacter vessel closure head will be inspectable.

Longitudinal and circunferential veld areas en the reactcr vessel interier sur-faces vill be inspectable.

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tiened above vill be established during the detailed design.

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The assembled rtacter ccciant systen vill te tested and inspected du-ing final nuclear unit ecnstruction and initial startup phases, as fellevs:

L.k.2.1 Eeacter Coclant Syste: Frecritical and "ct leak. Test _

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This test de=cnstrates satisfactcry prelitinary operaticn of the prescu-icer and itc individual ec penents. Spray valve adjustrente and heater centrol ad-Juctrents are tested.

L.L.2.3 Relief Syster Test In this test all relief valves are set and adjusted, and cpe'ating ;rcredares n .e.a

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Schedule for Capsule Re:cval in Unit 1 Equivalent Vessel Material Ecider Tube _s Exposure Time (Years) 1 10 2 .

20 3 30 L L5 This evaluation vill be accceplished by testing senples of the caterial frc:

the reacter vessel which are ecntained in the surveillance specinen capsules.

These capsules ccntain steel coupens frc plate, veld, and heat-affected ccne raterial used in fabricating the reacter vessel. Desiteters are placed with the Charpy V-notch impact specinens and tensile specimens. The desiteters vill perrit evaluation of flux as seen by the specinens and vessel vall. Tc prevent ccrresicn the specimens are encicsed in stainless steel sheaths.

The irradiated sa ples are tested to deter ine the caterial prcperties, such as tensile, impact, etc, and the irradiated NDTT which may be reasured in a nan-ner similar to the initial NDTT. These test results can be ec pared with the then-existing data on the effects of neutron flux and spectrum cn engineering materials.

The reasured neutron flux and NOTT may then be ec pared with the initial E;TT

% and the predicted UDTT shift tc coniter the progress of radiation-induced changes in the vessel raterials. As the end of reactor design life nears. a significant increase in reasured NDTT in excess of the predicted NITT shift could be investigated by reviewing the vessel stress analysis and cperating records. If necessary or required in accordance with the advanced kncvledge available at that time, the vessel transient licitations en pressure and ter-perature may be altered so that vessel stress limits, as stated in k.1.h.3 fer heatup and cooldown, are not exceeded.

L.5 CUALITY CONTROL h.5.1 CEERAL L.5.1.1 Dircnsional Inspactitn In-process and final dimensional inspections are made to insure that parts cf arse blies meet the drawinc recuirements, and an "as-tuilt" reccrd cf thece disensicas is kept fcr reference. A temperature-centrolled gauge rect is maintained to keep all reasuring equipment in prcper calibraticn, and perscnnel supervising this verk are trained in formal programs sponsored by gauge equip-cent =anufacturcrs.

L.5 1.2 Ncndert ructive Testing i

The primary purpose cf ncn^rstructive testing is to 1ccate, define . nni de-termine the sice of sterial cefects tc allev an evalustion cf defect acce;;1nce, rejection, cr repair.

ifUso m y;.:s q

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a. Fadiccr thy Eadisgraphy, including x-ray, high-voltage linear accelerater, or raf.icactive sources, will be used as applicable cc determine the acceptability of pressure integrity velds and other velds as speci-ficatiens require.

Radiography inccrperates the fclleving techniques:

(1) All velds are properly prepared by chipping and grinding valleys beteeen stringer beads so that radic~ orat.hs can be .trceerly in-terpreted.

(2) All radicgraphs are prepared and reviewed by qualified personnel.

(3) An c . 0S o- in. lead filter is used at the fil: to abscri " tread-bea: scatter" when using high-voltage equiptent (above 1 MeV).

(k) Fine grain er extra fine grain fil= is used fcr all exposures.

(5) Densities of radiographs are controlled by densiteteters.

(6) Double fil= technique is used en all garna-ray exposures as well as high-voltage execsures.

,j (7) Films are processed through an aute=atic processcr which has a centrolled replenishment, terperature, and process cycle.

(8) Energies are controlled so as to be in the optinu: range.

b. Ultrasonic Inspection Ultrasonics are used to examine pressure-integrity rav =aterial, and the bond between corrosion-resistant cladding and base material.

Ultrasenic inspettion $3 used as fc11cvs:

(1) In order to deteci latinatiens which lie parallel to the surface, plates are also inspceted by a lcngitudinsi (or normal) tea technique. The shear wave is scre effective in detecting de-fects criented perpendicular to the surface.

(2) All plates are 100 percent volumetrically inspected by ultrasonics using both normal and shear waves.

(3) Closure stud forgings vill be inspected.

(h) Personnel ccnducting ultracenic inspections are trained and qualified.

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. -s * :. n w e..e. c. a. . .~. e e. n.. a. A- -c . . a. n, .e .1 4. .r .:-.o ;

A.

y 4 , , 4 ,4

.g.* ;E ..e +.e- ...

. . r-f e . .e. .e-.

~ 4.

n

.4 Y 4 n. c. y g .. .4..- c + A 4 c e.. *f.e. 4.- F..

4. S "ep4 -- . ~"e." #. . - c . # ' ',. *e.,, # F. *. ..e a *e"- - . .

%. 4 A. y . o. r. . e. *.-.:= .. .

...-c e+.4* e p.3_..g i .e. co nw g e. c. +. -o., . .,7. e--e. e. e. ,7 4e r-.e to-A e e eC*.e .- _ .. .. - e . . .. . . . - - - - . . . . . ._

.w.. e e ,A s. w, A e e_ y a. n s. .~.a

. ~ - A .Ac n

-syu-.-me:+ 6=- w- 1sA5.~. .

g,. e . .. ,.e :.e..a.-  :

- n t. A-w af .e. a. +wv . e ..+. c.y c_ 4...c c ..a a 1 c e. .. 2. . . c . 1 4 .g ..e.

-r e.,. .e v. ,. n. . 4.

..- , e 7 4 r. 4 e A- .

y.).3.2 y ,. ,.-_A : . . ,- .c. . = =.,_E.. . .. ,_,+:.,. . .=. ,.. - , e _c e n. e . _

• . ' . " . " ~~**.*-.'#E., E ***~'E". C.

- 1#. #. 4. C 'c *. #. w".'. C.# E'1 '.**..I.'..e~~'.'. c..

'y.* C.# .. .

"w, 'w*'

a ++ . .vang ntwev -- . . y. v e.

^

1. r" "- _a.. *. * '. ' "w .* e # ^ S '.--.~y"w"*,."*.#. F. e.

. v.. . 'c*. Cw ^ .- A & #. e * * "_ a. *. '."~* ". 4. ^e E '. . . .

v..e E.. + r e c..4 .. ,- y. . v c.e.en.e .

res. C c+., 4 4. e oA g , +.,g..e. e. . -w 4. S ,,--.c._

_.-e.

..,77 e e. ... .: M. 4. .e w..

p C. +. %.. e ._. w .

1 a.-. .4. E . A u. . . w 4.... e .-.4 m . . . . . . .

+wa

.,u. "y".C"a.".'.4.*.'.'

y C. w #

• b..e ~.~*

. *. * *. . 4 al , E e.,. ".eae4.Ve#,

v *".#.

y e .~_ i. *. C e " .* .4 .. .# # ^ E*. # ^ *.. '. k.. c' *.

• k.

. . -. a

.#i ". . #. e *. .

• A *e ..w ar* -. ,%,.,'..E...

a^ --

g. _e *w *

"_ e C '.". E".. ..#

• a.' y*.* w^",, * .* . *. # e 5 C .# *".a.

.'. ~_s~ ' .e "#. c

• 15 .. # ".

• wb..a- .

'w ..". '* a ~. .'. e- '. E *~* . . ,' "y e .'

. ae . . "y.CE'2,.** ,

"w _c e w" . . .

C v. ^ *yw^ ".. .e. ".'. c E " e C # ".S ' .S '. e *.*. .

• ". *. .*. '. .". ^, .c -a sy4e

. . . . --nc_

7. we C C e.. S 4 ..-e. + S C r. ..

4 7c.pe p e. r. g- ..-4... .f.--.cAn. e. .e ... - c._

w..7A

g. p c.7 4.#. 4 c g..4.

w ,,,.. +. e . e. +. yA ._ . y. . . . . . .

,pe+ +c *

.g ,. b. . o *.

• v. g e . n e.*. e +gvp. . - c. g

.a e..e. #. b. .a-. c .o . g ,.. .

4pn.  : S. g . e e ow _

. - w - -

g-o ...3 .... 7

+ 4 ,. e. . +.o. Q w c1e+ag

+ +a g77 e.r e. 2 o . w . ..4..e . c. ..- e

v. cww. ,% g* n *.a. e..g ,g. . . .: 7 J. #. . n 3-..w y G. ww * - .. *- . ww m,2. . n.w. :. w. ye yve.. . , . ,A .og . g _.* a. y e.. c._ ... e.. e e. ., , _n._.,._....:..a.

% c. p~y3gyqA

_ . a .. f

.v .

- -. e e. , .e .w. m. b.. .

...e4-g e. 4 .w.7....w g e. .- .,. . a. s'w ma. g e . . -e w. 4- e.

cwe .b..,. e.a .e.

.,g .:. e i. p , 3.< e..n -o c 4. e.e..

f r., wwnA..p. e*

%. e y . wn -...w...

e *^ e .c ' c#. *.#..i .e .' . -

• '4 '^ "* ** C S *- *' r 1E' e S *O C' e _c '. *. " "v *o .# ". e *w e s *w . ..

C. C"%hie ^'i".6

. .~ *US14 ##- A

• a C E**

s o

m.oc..S41u e e+. _ . ,e..e.s...

p. . . : 7 4 . .. .

..... A.,.

hqep. -o.

• ar 7 ,

E 's. r e * .

O.... p e a....e

• p .=. .i. e

• 6v Y. w.w..**ig r.e g e. *. , .* g g #.

. . *wh. . -a =A7A. .. o.-~

• r3

.e c e. p s.

. . ..A .g..

. .m. e. n. (. U. .L.* s\ . .c.'. n-* ? .

v f h DN.J * -*0

..c0

i e.,'

. After cc pletion cf the qualification test program, production velding and inspecticn prccedures are implemented.

All velders are qualified er requalified, as necessary, in accordance with S&W and applicable code requirements. Each lot of velding electrodes and fluxes is tested and qualified bercre release to insure that required techanical prc-perties and as-deposite'd che ical properties can te net. Electrodes are iden-tified and issued only en an appreved request tc insure that the ccrrect ra-terials are used in each veld. All velding electrodes and fluxes are main-tained dry and free frc: contamination before use. Records are maintained and reviewed by velding engineers to insure that approved precedures and caterials are being used. Records are maintained for each veld joint and include the j velder's name, esseatial veld parameters, and electrode heat or lot number.

t 8

b.5.1.L Material Identificatien All plate cr other caterials are ;erranently identified. The identity is rain-tained throughout manufacture so that each piece can be located in the finished

vessels.

I h.5.2 REACTOR VESSEL

' The reacter vessel vill be designed, manufactured and tested in ac crdance with e

Secticn III of the ASME Code, 1966 editicn, including the Eunner Acienda cf 1965.

To insure ec pliance with the ccie requirements, reacter vessel manufacturing

' f'% and testin6 procedures shall be based upon the 3sW Quality Centrcl and quality A2 Assurance Prc6 rams which include the folleving require ents:

a. The heat nu:ber, chemical cc position, and techanical prcperties of all pressure boundary base caterials, as defined in AEME Secticn IX, vill be obtained and reccrded. A reccrd cf all pressure boun'_ary =aterial vill be raintained to =ake it poss-ible to relate each ec penent of the finished vessel tc the certificatien of caterial and fabrication histcry,

b. The pressure . ' ate caterial vill be SA-533, Grade 3, Class 1.

The fergings vill be ASTM A-5CE-6L Class 2, as rodified by ASME Code Case 1322 L. The internal cladding used fcr corresien resis-tance vill be austenitic stainless steel. The vessels a-a 4-e"'ated with a cc bination of metallic reflective and Icv-halide-centent eenventicnal cass insulating materials which are ec patible with the cceponent raterial.

c. Coupons for mechanical test specimens vill be recoved frc: "as-fabricated" caterials and veld precedure qualifications en sin-ilar =aterial vill be used to certify veld retal for "as-fabricated" caterials. The test specimens vill be subjected to a pcst-veld heat treatment equivalent to tha --an' ants that tka ,-c 's they represent vill receive in the etnpleted vessel. If tne test raterial interferes with perforring cperaticns of a part, such as

=ay cecur en heads, separate test caterial meeting Paragraph 5-313.1 g-)g

(_ cf the AEMI Code, Eecticn III, tsy te : sed.

L-27 GO' W' 33-  :

._=...=_-__...._,c z r / 0 0 .' A' ;

e,-- ,

d. I pact properties of all specimens rencved frc= "as-fatricated" carten and Icv alley steel, inclu: ling veld precedure qualificatiens

4. . a . . ~. . da . -. e v.' . *.. '. e . ar. . e.-y *.. .',' ,c L O c'. ce,.*..*..."1., . . T 3..v. *

..rmc d. a_ , ~ . .' . ' . '

_... +w. e e v"4. e..a....s c.' .%.a3.s,....

. *' " ."O, .ca...'~.. a

. . 1.

.T, acv. r. c_ ,4 ..

Tne te=perature at which a Charpy V-?;ctch test result of 30 ft-lb '

is cbtained vill te determined. The ?iD?I cf the vessel esterial cppesite the tore regien is expected to te in the range cf apprcxi-rately +10 F to +25 F. These Charpy V-Nctch impact tests are per_

fer ed en base =etal a.d en pressure vessel test plates,

e. The fc11cving delineates inspection of pressure beundary raterials in accordance vi h AS?G, Sectic: III, requirerent s for Class A vessels, or where the ecde gives a choice cf techniques, and states vku 4.cku c.' *eh. e a . . a. , '. aM. . a. 4 .. e. ,- a. e

• i c.,. . " 41.1

. k. a. - ~.. ' - +. e d ,

mA3.ic a2' ,

specific inspecticns required by EFa' are Else given:

-. . . . . ..e.ec~..

c. ~. v. . ~. 2.3,.

. , ,1...e. v41x, s. o. . . .. . . ecH...e=.v .

a. 4. .r.s , w + e 4 0..e k" .u . d e '

• a.

- y .- .e ..+. 1 .~. 3.

~4 "u

d i n a'. "a". e ' r.. e. -, e c + 4 .. +k...~."-3... ~k the entire vcluse in acecrdance with AS'E Secticn III,

.Faracranh N-321.

One hundred percent shear vave insrected through the e .. 1

. e ' a_. . . ..

2. In the case of het forced plate, cne additicnal ultracenic in-

) spection vill te perfcr ed upcn ec pletic cf the final fer -

ing operatica and any heat treatment espicyed except stress relief.

3. All pressure beundary forgings vill be 1005 ultrasenically in-

• ed t? -u "*.

• k.a. *

.e.,ea. c"-3' .. e ".*. i . e ". c 1" e .'. a . ~. d a n. e "- t' *. k.

. . . . n cv. ._r ,

Section III, Paragraph N-322.

L. C1a,d d1' . .3~ "4'_' -_ k a. 4

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

C1ad d. .4 ..e c.

. '.k..e . e =.-+.c

. . '. ' ar.e a e. e a.' e"..'a

- a= ".i.

.. be "'+w.=-

s e e.t. . C a.1 1.. .i .e.,- a n +. e A. . . eC.

. 1 twCv ,., , s v.r + %.. a.

,. ..C*.e-.C.

r

.  % ~r.h % C .-

. .A and defect.

n,,

- , , . 4 4 4 ., e--s. 4 3 , to. ,,,..ge-,4--.

c..v..o. .,...---s.. . . . . ~.---... .i .n s, ~n~ *. a. d ~c - a ex. e '..^ '.r.

4

.c . . . .

S c ..d 4..3~ >=.."aa... -l . a". a ... d. S aca.

. .- .. = 'v e .4 a _' .

5 Tne area of the flange seal surfaces en which the double seals seat vill te liquid penetrant inspected per ASIG Code, Secticn III, Appendix IX, Faragraph IX-360, after the hydrcstatic test.

c. ->e s .. .,1n,,.

....~4e e . .. y. +.

- .4 ..

. . a. 1 . ~~,e '.'e.ns c' ,-.=.se"

. . . e bc"..d-a-". . .

Caterials to AS'E Code, Secticn III, Class A ressels, standards

"- . ' . ' .' b e C c . '." . *wd e -m .i..' b.e s' c. a. C.# .' ak. .. ' -..

_ . =- . ' r . as fc~w.e- . ..

a

\_.

tsa 000G2 .

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

c w =,:0 si -.: n

C ' E ^ # .. * '. .-- . *. .' .1 ** w. A'=.~"..'.*.*..~E..*.

-; y *ee*e#-

& . *. e . .~ ~ ~ e s e -..s. *'*e#. .

C e. -v. .. + ., g.A.. . .7 e.. . g1 1 r.g., v.oeg .. e..+., a 4-..g. .4.71.

. s a. ..s :. . a + 4. a.

. ,c..+.,

.a n. .' a. .. t e. e, .. a. n. c A. E r. .. a. . g.. . g .e. n. .'e. e..A-

- .. c _.y- a. .

r.11. g . ,. e .e s <.> , e ,. . ., ,2 , v.4 -1.,

. . v.a w e 4 ...v.a. . . e g..

. . a. ..4 .- ~ .

ga..... 4-,e c. .

A g. . .., _g- .

a. ...a. *. e. e. +. 4 . ..c.; e *, e d E#* e * '".*i .

d" ; 5

• E*. i '

• 9 5 *. . .

'.%.., e X.. e.,... s.1 . . . 5. . ,. c-. . E n. e C sP *...w a. <-

• e e g a .' , 4 e..w n ' . . -A. .'- .e. g. .=.e .? A- .ca. p. S, vd l.'

. k a. . ~e-- g .. =. *4

. a C

• yc ." +. 4 C ' a. # ".

.. .cy."

• C * *. - # ..0..#

• e " "#

.# .# .". .S ' .".; - ." w' *. a a. e*

• A A A.t..* a. c.=.o- .1 f e. . .e ; - a. e. s. $ .. e. e. e ,e p .mye. g, Es. .t e. _. e. . p e. A. ,e y .t a. .. w r f a 5 .4 e. g.. , . e .n .e..

. e , c. -.A- .A .a .es t g. . e.

a .g... 4 , a. e..4 ..$~ ...411 v.g c-.A.. .. . .e. A E, . , ,

e n.,g - . a A. . . .c a. a. +. . s. v. g =c -

r~.'.r ~ . ~. . ~ .e g.=.z. ., , . c. ..* c. .

.Ie . 5 . .. .

u s o.

h.. .e.,,.. .r. 6 w . .

'."'v. . a.e*cc

e. ~ g e .v. a. = -e

• w. * * * *..' . ' . ' a. -A.c e 4 g. . e. c A , -,e.".*#c~~.~'.**A.-

-s. -. n '. A- **=.c'*=.* ...

u* .. .

1. • %~

-.~".'--*..'n. **"

. 44.* v. c e w.

. . 4. w.e4 . T. T. ' C.# '.'".A. . r'. M. L^ v* .#. e .D..".#

. . ... **#'.' k. *. e ".k ,#. a *. * *. #- **

. .w ' ' . " . * .v44w...

.# ^ 7 ' * * * # ". g~. . " w. " " . . -

a

.a5 e , . .w e s.a. t ..e + a 5 +. 4..e..- C' . ...,. 4 .s. e -- c .e. . . r- . E n.

... ... , e .

E. O1o+e

. .. .. .411 Ka.w 17.,&go . . . . 4. .w . C E.1 1.. ,.

4

..y e

a.a+e-.

w.

A

>..  : n. g v. g. .. . .a. a. . . v. .. a.

n g.-AAt g.e2-A %gea. g.. g :. .a.n. . . ' '

. ca-.. .g- . .... v a. . . -n .+. ,.- *. c. C '. . .-

1. ^ =- .' .' "e d e.5.yw.

.. aa+gd.

J

~

f g

1

%s i - .

0 L. $T - ..> 3

--g .

-" .a s.

1 I

i

c. Welds vill be inspected radicgraphically cr ultrasenically, as applicable.

d. Weld depcsited cladding vill be liquid-penetrant examined.

i e. m1 s o. ... c .+ u a . o. .--

e w. ,e. . o. .,a.e. v.4 .,1 w. s, ,4,m.44-,.,.+.-,+.

.. e. .-- exa 4.-.a.. ..

} .

L.e - .1. c:.._ .r.e

1. T . _ .:.

I t

j The pressurizer vessel vill te designed, manufactured asi tested in acccrdance with Sectic: III cf the ASME Cede. It vill be subjected to the folleving non-destructive testing durine zanufacture:. .

i

, a. Plate vill be ultrasenically inspected.

4

! >..  : n ya . .~. ,. . .- -

.. , , s a a : c 3 ~4 s. e .c a. -c.e i.al ".il' k a. . '. +. . a ..e r . . '. r a .'..' ~. 4 .-

.e-,..

y..... a.

f c. Welds vill be inspected radicgraphically or ultrasenically, as I at .r licable.

d. Weld deposited cladding vill be inspected by the liquid penetrant

, . ,. ..u.. a . .

Ef n. C,i C5 C n0. v .L.n.a

.-, -. 1-. ..nri r a.e u 1

• *.$.ej O e.s.

u.

.. ac*..-.w -..'a-+ --. y -ir.4 e~ vil.l c ^ ..#. c .-- +e o +.k..a. Us~n' ST. C.^ d.a. c a. c+. 4 --

~ .

February 1963 plus June 1963 errata plus applicable code cases and, during

. - . . 3^1.T -

1 fabricatien, vill be subjected to 'the following ncndestructive testing: .

l.

a. Bond between cladding and base =aterial vill te ultrasenically in-sy ected.

4

b. Welds vill be insrected radic s ra-bically v . and t.v tarnetic particle examination, as applicable.

c. Weld deposited cladding vill be liquid-penetrant exasized.

L. . .c -. -

. .:_n i n CD.,7_v..,

s. i r.,,

. v. , Cr, e. I...g

. .. u The purp casings vill be ranufactured in accordance vitF Section III of the ASMI Ccde, where applicable, and vill te subjected to the fc11cvir.g ncndestructive tests: .

a. The rcugh casing vill be inspected radicgraphic.11y.

,,i

b. The finished casing vill be liquid-penetrant exanined.

c. Welds vill be ins eected radicera:hically . . . and t.v li.uid penetrant

=etheds.

D 1

ms 00034

,. -c ... . 2. .. e n-cv .--. - . . . .a . . ,

3 ./ .

- _ . ' .' c

^* . 6 t. .r 77".r'eP

- w LO (1) Pcrse, L., S.a.aetr.* Ve  :

. . _ 4. 4. c - 4. . . r_ ". . e ". e . ,

_e

.c. a. '. D a. c _4 .. C .- . .e

. . 4 4. c_ . .' . . e-ASME Fa;er ?;c. 63-WA-100.

(2) Fellini, W. S. and Pu:ak, P. F., Fractu e Analysis Diagran Fre-cedures for the Fracture-Safe Engineering Design cf Steel Structres, Welding Ee earch Ccuncil Eulletin EE, ':ay 1963 t (3) Eotertsen, T. S., Prcragatien of Erittle Tracture in Steel, Journal of Iron and Steel Institute, Vel =e 175, December 1953 (L) Kihara, E. and Masutichi, K., Effects of Peridual Stress en Erittle Fracture, Welding Jcurnal, Vc1=e 35, April 1959 (5) Miller, E. C. , The Integrity cf Feactcr Fressre Ves sels , CF !L-::F:C-15, May 1966.

! (6) Topical Report E&W - 10,018 I,

i Table L-1 i

i Tabulatien of Eeacter Ccolant Syster Pressure Settines 1 Ite- Fressure, Psig Design Pressure 2,500

[ .

s' Operating Pressure 2,185

]

Code Relief Valves 2,500 Pilot-Act. Felief Valve 2,300 High-Fressure Trip 2,350 Ei Eh-Pressure Alam 2,255

.ev L ,rressure a.,a =

n c,1,v

,g-Lev-Pressure Trip 2,050 9

t.

f % 4) %

• O$- r L -29  :- _.

.-- .. . ..c...

b -e

• V, ,

, - m y

Table L-2

e . - . u . v-,. a. . . , ,,-.

Parameter Value Tctal Solids, Max. (Including tissolved and Undisselved Eut Excluding E 30 and :05), pp= 1.0 a 3 Seren, pp= See Figure 3-1

.~u =- tv , ---  :- 6

-u- -- e+ ~7 t i. L.C. y .3 pH a: 560 F (Calculated) 6.1-7.6 d2 0.01 (Max. ), pp=

C1 (Max. ), pp= 0.1

, s

u. . --,ca cc/., .es--O Eydrazine (Eequired During Shutdcun), pp: 25 e

k k

v e

-v s

i l

I I

i 1

Table L-3

, Reactcr Vessel Design Data Iter Pata 1

Design / Operating Fressure, psig 2,500/2,185 Hydretest Pressure (Cold), psig 3, 25 Design / Operating Tenperature, F 650/600 Overall Height of Vessel an:1 C1csure I Eead, ft-in.

! 37 L 1

Straight Shell Thickness, in. 8-7/16 Water Velute, ft" L,150

,! Thickness of Insulation, in. 3 4

I;urber of Heacter Clcsure Head Studs 60 riange ID, in. 165 Shell ID, in. 171

, Inlet Nozzle ID, in. 28 e

i Outlet Nozzle ID, in. 36 Ccre Flooding Water I;c :le ID, in. 11-1/2 Tatle L L Fressurizer Design Eata Ite: Data Design / Operating Fressure, psig 2,500/2,155 Hydrotest Fressure (Cold), prig 3,125 Design / Operating Temperature, F 670/650 o

Fortal Water Volume, ft" 800

!;cr:al Steer Vclure, ft 700 Surge Line :::::le Diameter, in. 10 N

1 U Cterall Heid.t, f:-in. L! -O s -

000'7 so-[,s t.

Table k-5 Stear Generater Design Cata Iter :nta per tri Design Fressure (Eeacter Ccclar.t/Stea ), psig 2,5:0/1,c50 Eydrctest Fressure (Tube Side-Cold, Reactcr Coclant), psis 3,125 Lesign Temperature (Reacter Coolant /Stea:), F 650/600 Reacter Coclant Flev, ib/nr 65.66 x 10 Heat Transferred, Etu/hr L.21 x 109 Steam Ccnditions at Fated Lead, Cutlet ::c::les:

Stes: Flcw, lb/hr 3.30 x 10 Steam Te:perature, F 57C (35 F Superheat)

Stean Pressure, psig 910 Feed ater Te:perature, F L55 Overall Height, ft-in. 73 1/2 Shell CD, in. IL7-1/L 2

Reactor Ccolant Water Voluxe, ft' 2,030

{ {<Y' ~ F.1 L.:= -

~~.

i t

1

- - s, - g

.a..e 4, .

Staa Ger.erater Feedvater Oualit.

. a-

. . L.. . . . .

p ..o..

1 n . n 1, .

r. t y EX . ) . r . :, r . gs \ . ... . . y Disselved 02 (M***)' EE= C' OI

c. .e 0 ( v. . x . s> , ,,...

. . 0 . n.c-d Fe (Max.), pp: 0.01

. e.. a,. c. e, : e. e

. .. . (.v. , x.i, y,_ n.oc v %

Cu (Max. ), ;;: C.C 2 u.a,.

. . ,e3 .

...a . ~ .r ii .e . a. <. .< C ...c. . t n a. 5 1 < . 5.

.. ..a. e c,.. . , e c e. . ...... .. n ~

a . .a.'. .*.=.

. .. #e_

4-.'.a+.'.,,...

__ ... .e.*e . , a. e .^...~u.ld. . $ a.

.o .ct., _n . c. ,. , 4 .....s 4 . . .,. .. .w .,. ,.a . . e ., .e e c .. ..<. .., .. ,.. .. .. s .

v e a...4 + - C . e- = . ". '.-..-- ...a

_. 4.a.i.-..

.. . . . ^..~ '.a. a". ~..' .' a. #. 4. . .

,.....a

,. . a . .. c. a . a. . e .- a,. . a . . e e . .' .' s 5..a. ' e... .e. . a. .e. . .e .e.. i -

S,, ,. . 4 . . , t .

.. . . . . . r. . . . . .

s Lead - Lead ecr.tamir.aticn of the feedvater cust te avoided. .

W s (t,p')

e 5; _ < ..

ud

t. ..

,-- g - ,- ,-

-- Table b-l Peacter Coolant Puru Desien Data Iten Lata ter Unit

.unber cf Furps (

2esign Pressure, psig 2 ,5^;0 Eydrotest Pressure (Cold), psig 3,125 Design Temperature, F 650 Operating S.:eed (Nctinal), rp 1,180 Fur;ei Fluid Ter;erature, F 60 te 5S0 Zevelcped Eest, ft 370 Capacity, sp: 88,000 Fyiraulic Efficiency, % .

$6 Seal Water Injecticn (Max. ), sp: 60

) Seal Water Eeturn (Max. ), sp: SS Purp Net:1e ID, in. 28 e ._

.. . .. e , a., ., Lr .. <.+. =. e <-u 3.. . , .e.. 2L, Water Volute, ft 95 "ctor Statcr Frs=e Dia ezer, ft 8 Furp-Motcr Mccent of Inertia, lb-ft 70,000

,or Data:

Tyr.e Squirrel-Cage ~nduction> Sinrle Sr.eed Voltage 6,600 Phase 3 Frequency, cps 60 Starting Acrcss-the-Line Input (Ect Reactor Coolan ), kW 5,600 Input (Oc1d Reacter Cool'.nt), kW 7,h00 00eri0

%~j4 L

.i 4

} Table L-S

,  : e . . ... c . , , . P : 4 . -...._...e:.

.-.a

..,.-_ ,a.,

t 1

1 t. . e. = . *. c .* , * ' a. '. 0' .. ..,' .e- TD , T

^*

c.

I

..., m . .. c .- Ou . 2 e '. I. .' ,.' . . 3- T. D , .4. .. _' #.

F- c E

. u-i . e. m u. e.a. .2 4,. . 3 ..e~ , i *. . . .O c '-.. .' - ' O i-p e..e . 4 .. /c, e... . .. 4. .e- :. . e s s' ~ e , , .=.e4 - 2 , c,0 0 / e^ ,. ' " 5

, Eydrctest Press re (Cold), psig 3,125

..o.c.e..jn

. _- ,.. 3.....e . ..u. e, .

_ .,..c. . c. e,v.f r. ._

e-

.n. a c. ie. - . / *ef- a.. a . .' .. e . e . ,- <= a* . a.

(.p ...s...

. - . . 4.e..

e...

-3

. ..4,.,),

..u t ;40/c.pn

. s

,t a

I ..

n a.. e w 7, , , ._ ,,

e.--, .c.. 3 , ,,

! 2,910

". ak.la. L-0 .

N

.) ..-. .. .e ... C, .. ,o..e

( O.n

. . . e A

c. .a

... T.

. a A _n e e. .:s

r. .e.

. . : . . e a.

._c.

Transient Descrittien esien Cycles Actual Cycles

1. r.ea... ,

. , 70 .-- 579 .:, a..a . -C . c,

. . a.,

579 to 70 F LEO e0 2_ . .v.....,

. , . ch3 +~- 579 F,

, , a.' Cm'o'd 2 '1,o +- O 5LO 7 .

' ^

2' , *'

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m

( .3 )--lu v -15r) s ..:,n-- O y,u. .

e n L. e . a. ,- _uw.

7-e*

4.e~.rT...a.ase (10')

. s -? , " .w." " ' :^^-

-,s 5 Step Unicading Decrease (10%) 2,000 1,:^3

6. Step Lead Reducticn to Auxiliary Lead e ,

cm

( .~ cu-e p ,

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.-. Lm .

a s E. iscellaneous Transients 10 5

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-- L _..,. . . .

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... (ee-. cw,e L.o,)

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~

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g v / v - . r . -. v. o c .-

t t ,; . O.2 .. , T C

. . s' d .

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• t 24 -

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. , . . .-.. . , ., * .* c a '. .= .. :... e ~w *...-. .. i . e. r.c ' ..e a . .e .

(cTe..=.4...c....-4.,,.

.vo . . v.o,- ... .... .e .z. .e .e ,4 2 4 , .

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1 J

l I

g TABLE 4 - lla J l s '

ASMI CODE CASES I Reactor Vessel 1332-4 1359-1 1441-1 Stes: Generator 1332-4 1401 1407 1492 i

Preosurizer 29 1332-4 1359-1 g) 1355-2

, 1441-1

(

Feacter Coolant Piping No ASMI Code Cases are used, as such, for the R.C. Piping, since this ccrpenent is based en ASSl B31.7, and it is considered incensistent to use an ASME Code Case to arend a docurent prepared by ancther certittee. hevever, the equivalent of the technical content of ASME Code Case 1401-1 for half-bcad veld repair of cladding has been incorporates in specificatiens fcr this piping as a recuiretent.

ANSI E31 Code Cases 69 and 79 are used. The technical ccatent of 1459-1 for half-reld repair cf tase cetal is incorporated inte the Spec;ficatien.

Ccre Floo ing Tank The equivalent of the technical content of AS!'I Code Case 2 23E-3 for inspection g

has been incorporated inte the Specification.

x- -

?

=<=drer: nc. 29 i-3C2 1.' ~ 5

Y e

i 9

i i

Table L-12

,s... < 1e_

m,<. C . ..-e .-., .n_

G C.-.-..._..-

. ection

!5terial 1

i

?.e a e:c r Ve s e =- .' .;.- a- e- -e .. _, . , -,e

- c.

-n-;;;,

ara.e a e, C ,' c e e

/+)

s r..

j P e e -m. c.,. e i s._ ng s A-50 -: , Classes 1 and 2 (Ccie Case 1332-L)

, p ; c ---

,<. ,' , -e.,',-

C1a .~ . .4r- s

' .- e-an a T..

_ . . e. ..

. . .. _, _e.

e-ve.. - s Ln, m.....  : n' a2

) ,-,._e.,._., .... . c.,,. . ,. e

. , . - - j,

...--c....,.

i s-- ...e.e n-: :-: , ,, -. asses .1 sn: c i.

(,v c .e _1se ,1 :- 3/

- _ -i 1.. ----_

. e :, o'

, Cv-...,1..w.. .,.avo < v e .v.a-m-n.,.<

e- In,-- -.... .- ==16~r e*-"- +

i and In-Core -""-'e"'

Fenetrations

ear Generater Pressure Pla = 'c.^ , - c- ' g.,a o

' o SA-533. Crade 3, Class 1

-tressuret ort i r*- e ~

A ':^' ' Cnes 1--

(C :e Cise 1332 L) .

18-E Stainless Steel C1 add .. **~

.% .c.

, Jeads C'_ _a d d.4 . .e- 'or Pubesheets Ni-Cr-Fe c .,':

v.,

m.G.ee.k--

--- lee- n_:,

-  : _t . C oc.e .' (",

-- ,u..- 'a Case ,___  ;;: - '); en,-lCc, G-="= --- _. 2.1. _01_c', '.,.c ,-

^?v.

.-re s surine r Shell, Heads, ", E a- v"l.. ' e-"^- .. , -^:~---

c-t

"=-.-a ,

-n r late

-ee.eL, e t o. e- n6s A-503-f . Class 1 C^.a. Ceea 1._,42 L

R &

Cladding It-: S:linless Steel

.* .. e = np o++ e . . ~

..F.<-e -., 'sC- I

< Q-- <c

.e m-

• g , , = 1 C, _1y 1=enti=a1 t0 bn* vie dr ce 3, as ,:21.p :ee:

.va1A'"*b

- re

, '- . . u _ i_ e m"s e c_ _10,0 g

%.e d

O% -

f A

s 4

Table L-12 (Contd)

) - '.c.+...2,,

t.: .;:ne nt cecticn ~.

Internal Plate SA-2LO, Tv.-e e ;Ca I

re..e...,-...-:<.. v e.-u.e .e .n-::c, . . .. e:,

i I..c::les A-500-o , Class 1 (Code Case 1332 L); EA-lC6, G-a'a 2- cn' ^12, 5.;e.a .sri

,1 F.eacter C:Olant 23 in. and 36 in. SA-516, Grade 70 (Elt:.:s) l

. A-loc,, Grade C (St raights )

r-

-,F.,ns i

1. 00 . e_. ' . .'. a. c e. .em+ e ,.

1 C'. add'- r.

10 4.

s...

, >r-

r. -- a: , araae s:-:,6 v_.:..cc.e) e l 7'se=- _,16 ( c' .-a 4 s~'..+.

2 i

A :t6, e)

!ictries A-503-6L, Class 1 (Cede Case 1332 L ); A-336, Class F5y A Co

,sw-, m.;.y-- . E s:1s o'

}

1 i

O s,

4

?

4 e

1 1

I 1

h a

f.\

m.

ti d s.

9 4e z?

i i

.i..,.

9y Table L-13 Eeferences f .; Firure L L - Increase in Trarriticn Terrerature Lue te Irradiation Effects 1or A_;;2E Stee2 Neutron Eef Temp I.x po su re , I;_TT ,

I.'o . Eeference Material Type F n /cr2 (>1 gey) p 1 AS.'E Paper All Steels Max. Curve fc- 550 Data No. 63 'n'A-100 (Figure 1).

2 ASTM-ST 350, A3023 Plate Trend Cu.ve fer 550 Data a

,Cce.

y ss

,c

r

r. . .=m  :.... .,s- - c,

..- v.c.n,  :

. ..v .c_; m.-e . -

y50 "x 1.^^~ e-p 12.

. s e L A3023 Flate 6 x 10 ASTM,STP n-3L1, 550 85(*)

p cec.

5 AST':-STF 3'.1, met A3023 Plate 550 6 x 101C" 100

,,cc..

e 6 AST'-S ? 3L1, 10 A3023 1 5 x 10 '

Plate 550 130 (" '>

p 226.

7 AST'i-STF 3L1, A3023 Plate 550 1 5 x 10 19 1tg p 226.

8 10 Quarterly Report A3023 Plate 550 3 x 10 ' 120 of P.m gress,

" Irradiation Ef-fects en Reactor Structural Mate-rials,"11-1-6k/

1-31-65 9 Quarterly Report A3023 9 Plate 550 3 x 10 135 on ?rCEresS,

" Irradiation Ef-fects c: Eeactor

St

actural Mate-ria.ls,"11-1-6h/

1-31-65 j (a )T.-ansverse s;eci . ens.

__:c 00016

D Table L-13 (contd)

eu t rc r.

Ref Temp Ex;oeure, :Z'r ,

I;c . Reference  ?'.at e rial Type F  ?;/cr2 (31 gey) y 10 '

10 Q;arterly Report A3C23 Plate 550 3 x 10 ' 1LO of Prcgress,

" Irradiation Ef-fects or Reactor -

Structural Mate-riais," 11-1-6L/

1-31-65 10 11 Q;arterly Eeport /.3023 Plate 550 3 x 10 ' 170 cf Prcgress,

" .24g4- --,_

fects on Beactor Structural Mate-rials," 11-1-6L/

1-31-65 9

12 Qaarterly Report A3023 Plate 550 3 x 10 2C; of Progress,

" Irradiation E'-

r

! ) fects en Beactor Stn2ctural Mate-rials," 11-1-6k/ "

l-31-65 .

Weld lb 13 Weldire Research A3023 500 5x10 70 Supplement, Vol to 27, I;o. 10, Oct 575 1962, p h65-S.

,A 1h Welding Research A3023 Weld 500 5 x 10^" 50 Supplement, Vol to 27, I;o.10, Oct 575 1962, p L65-S. ,

15 Welding Eesearch Weld lb A3C23 500 5 x lo 37 Eupplement, Vol to 27, :;c . 10, Oc . 575 1962, p L65-S.

16 Welding Research A3023 Weld 500 5 x 10 25 Supplement, Vol to 27, I;c. 10, Oct 575 1962, p L65-S.

e- 0

s g!I s e,gt- ,.

= e-<+- e -(4 c'4 J Io r

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l

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ss N*h 4, 't EF'

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r - h- 1 .,t f

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4 f!E A C T O R C O O L A N T S Y S T E t' - ' F :. A N G E !' E!; T - E L E V '. i ! O N 00049 F i gu r e 4-2

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

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e e

l q'l

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12' - 5E 4-32

+

11 _7' "q p-S s-REACTOR CODLINT SYSTE'l * : E i N G E " E .N T - F L I N 00050 ..

t\[UTE 4.:

~

l i

,o _

l 1 2 4 6 1 2 4 6 j 500 , , , . . . . . . . .

1 2 4 6

. ,..,,,c,., . ,,,,,,,,,,,

M 2

en 400 c Maximum Curve 1

i

=

for 550 F Data ' '

/ '

- (See Reference 1.) /

2 300 a.

- /

c A

1 *

/ L Trend Corve 3<

h ~

- 200 12 * / 5 1

F Da ta -

m

• 11. / ( S e e R e f e r e n c e 2.)

.1

/

/g10 l C/* / 6*

i U 100 5

- 11

• 3

+4

/

o u

2 14+.

+-

/

j I B f '5 0 i ' ,'n' '

, . - ' i ,,'i' >>>> >

> ,,,,,i,,,,

!I 10 18 10 19 1020 102 )

In t egr a ted Heu tr on E xposur e (E> Me v), n/cm2 l

l Notes:

1.

All data is for 30 ft-Ib "fix". ,

2.

Hunbers on Curves indicate References in Table 4-13 m.

L./

>^

y.

Hit EUCTltlTT iRINSlilCN IErTERA'URE IECREASE iERSu5 INiECEliED NEUIRON EIPOSURE FOR A302s STEEL 0005.1 r; tere 4 4

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

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

i l

< O._

a

20'- 9*

f

( ACROS$ h0ZILE FACE 31 4

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4 ,

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+ ,

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/

g h0ZZLE SF E:Y N C Z Z LE' ,, N x

(. -Q r : VFL' L ;%E %; Z Z LE

.,h LEVEL SENSING N;ZZLE

.; (i! F I C A L GF 3) l, 'k (s g' I

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- SUD SE its * *ITF S H O C F. SHIELD FL C' KCL ES SUF.GE L INE NS ZZL E t.-~

v

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figure 4_E

INLET A,

1 P

O O SUPERHEAT REGION AT FATED LCiD n &

{

l EcrtE e

FIL 191LihG g FEGION AT d STE: OUTL ET Rt.TED LOAD RYP Ac5 STE A?' FOR

__kP{C m ,

3 0 0 7 r:~" -~

FEE ATER HEATi'3 I I F E E D a !.T E R INLET Tf i i 00aNCOMER

,[ S E C T 10','

NUCLEATE EDILING REGION - SHELL AT RATED LOAD I

REACTCR CCOLANT OUTLET q +i E y ,

e e..-a . - c, a -

'J l E ". e ! e P. ~.7, v

i i.$c F.*r

.c4 v 't 2 j j[. I[Ui6 -h

r,u 100 100 90 Superh' eat , Oc.nccrer ,/ 90 level y SO ' #

- S -

\l I~D  %

70

/

/ f,\

f

/ 70 f E o

60 t

/

/

[ 50 "

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