ML19309C542

Chapter 4 to TMI-1 PSAR, Rcs. Includes Revisions 1-11

ML19309C542
Person / Time
Site:	Three Mile Island
Issue date:	05/01/1967
From:	JERSEY CENTRAL POWER & LIGHT CO., METROPOLITAN EDISON CO.
To:
References
NUDOCS 8004080734
Download: ML19309C542 (56)

Text

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O OM.LE OF CONTENTS Sectica Psge h REACTOR COOLAITT SYSTEM k-1 4.1 DESIGN EASES h-1 4.1.1 FERFORMANCE 03.7ECTIVES 4-1 4.1.2 DESIGN CHARACTERISTICS 4-1 4.1.2.1 Design Pressurs 4-1 4.1.2.2 Design Temperature 4-2 4.1.2 3 Reaction Loads 4-2 4.1.2.4 Seismic M ais 4-2 4 k.l.2 5 Cyclic k ais 4-2 k.1.2.6 Water Chemistry 4-2 O

V 4.1 3 EXPECTED OPERA"'ING CONDITIONS 4-2 4.1.4 SERVICE LIFE k-3 k.1.4.1 Material Radiation Damage k-3 k.1.k.2 ' Nuclear Unit Operational Ther:nal Cycles h-3

.l.4 3 Operating Procedures 4h L.l.4.4 Quality Manufacture 4-3 k.1 5 CODES AND CLASSITICATIONS 4-6 4.2 SYSTEM DESCRIPTION AND OPERATION k-6 h.2.1 GENERAL DESCRIPTION 4-6 h.2.2 MAJOR CCMPC3EN"'S 4-6 k.2.2.1 Reacter Yessel 4-6 k.2.2.2 Pres.vri::sr E-7 k.2.2 3 Steam Gensrator E-8 4.2.2.k Res: or Coolant P.:=ps ,-11 L.2.2 5 Reactor C nlant Piping k-11

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O CONTENTS (Cent'd)

Section h

h.2.3 FRESSURE-RELIEVING DEVICES h-12 k.2.4 ENVIRONMENTAL PROTECTION h-12 h.2.5 MATERIALS OF CONSTRUCTION k-12 k.2.6 MAXIMUM HEATING AND COOLING RATES h-lh h.2.T LEAK DETECTION h-lk h.3 SYSTEM DESIGN EVALUATION h-16 h.3.1 SAFETY FACTORS h-16 h.3.1.1 Pressure Vessel Safety h-16 h.3.1.2 Piping k-21 h.3.1.3 Steam Generater h-21 h.3.2 RELIANCE ON INTERCONNECTED SYSTEMS k-23 O k.3.3 SYSTEM INTEGRITY h-23 k.3.k PRESSURE RELIEF k-23 h.3.5 REDUNDANCY h-2k k.3.6 SAFETY ANALYSIS h-2k h.3.T OPERATIONAL LIMITS h-2k h.h TESTS AND INSPECTIONS h-25 k.k.1 COMPONENT IN-SERVICE INSPECTION h-25 h.k.l.1 Reactor Vescel k-25 1 h.h.1.2 Pressurizer h-25a h.k.2.3 Steon Generster h-25a h.h.1.h Reactor Ceolant Pu=ts b-25a k.h.l.5 Piping h-25a h.k l.6 DJssimilar Metal and Representative Welds h-25a 0000 317 h-ii (Revised T-21-67)

O CONTENTS (Cont'd)

Section g h.h.2 REACTOR COOLANT SYSTEM TESTS AND 1 INSPECTIONS k-25b h.h.2.1 Reactor Coolant System Precritical and Het Leak Test h-25b h.k.2.2 Pressurizing System Precritical Operational Test k-26 4.h.2.3 Pressurizer Surge Piping Temperature Gradient Test k-26 h.h.2.h Relief System Test h-26 h.h.2.5 Statien Power Startuu Test h-26 k.h.2.6 Station Pever Heat Balance L-26 k.h.2.7 Station Power Shutdown Test h-26 h.h.3 MATERIAL IRRADIATION SURVEILLANCE h-26 h,5 REFERENCES h-28 l

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O LIST OF TABLES (At rear of Section)

Table No. Title Page 4-1 Tabulation of Reactor Coolant System Pressure Settings 4-29 4-2 Reactor Vessel Design Data 4-29 k-3 Pressurizer Design Dr.ta 4-30 4k Steam Generator Design Data 4-30 4-5 Reactor Coolant Pump Design Data 4-31 4-6 Reactor Coolant Piping Design Data 4-32 4-7 Transient Cycles 4-32 4-8 Design Transient Cycles 4-33 4-9 Reactor Coolant System Cedes and Classifications 4-33 O 4-10 Materials of Construction 4-34 4-11 References for Figure 4 Increase in Transition Teaperature Due to Irradiation Effects for A3C2B Steel h-35

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O LIST OF FIGURES (At rear of Section)

Figure No. Title b-1 Reactor Coolant System 4-2 Reactor Coolant System Arrangement-Elevation k-3 Reactor Coolant System Arrangement-Plan 4-4 Nil-Ductility Transition Temperature Increase Versus Integrated Neutron Exposure for A3023 Steel h-5 Reactor Vessel 4-6 Pressurizer 4-7 Steam Generator 4-8 Steam Generator Heating Regions k-9 Steam Generator Heating Surface and Downcomer Level Versus Power 4-10 Steam Generator Temperatures 4-11 Reactor Coolant Pump h-12 Predicted N!yIT Shift Versus Reactor Vessel Irradiation 1

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O 4 REACTCR COOLANT SYSTEM 4.1 DESIGN EASES The reactor coolant system consists of the reactor vessel, coolant pumps, steam generators, pressurizer, and interconnecting piping. The functional relationship between coolant system components is shown in Figure k-1.

The coolant system physical arrangement is shown in Figures 4-2 and 4-3 The reactor coolant system is designed in accordance with the following codes:

Piping and Valves - ASA331.1-1955 (Pressure Piping) including

- nuclear cases.

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

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

Pres 3urizer - ASME Boiler and Pressure vessel Code,Section III, Nuclear Vessels.

Reactor Vessel - ASME Boiler and Pressure Vessel Code,Section III,

( Nuclear vessels.

Welding Qualifications - ASME Boiler and Pressure Vessel Code,Section IX.

To assist in the review of the system drawings, a standard set of symbols and abbreviations has been used and is summarized in Figure 9-1.

4.1.1 PERFORMANCE OBJECTIVES The reactor coolant system 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 addition to serving as a heat transport medium, the coolant also serves as a neutron moderator and reflector, and as a solvent for the soluble poison (boric acid) utilized in chemical shir nactivity control.

As the coolant energy and rsdioactive material container, the reactor coolant. system is designed to maintain its integrity under all operating conditions. While performing this function, the system serves the safe-guard objective of preventing the release to the reactor building of any fission products that escape the primary barrier, the core cladding.

4.1.2 DESIGN CHARACTERISTICS k.l.2.1 Design Pressure O4 The reactor coolant system design, operati2g, and control set point pres-sures are listed in Table 4-1. The design pressure allows for operating ,

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transient pressure changes. The selected design =argin considers core thermal lag, coolant transport ti=es and pressure drops, instr.:=entation and control response characteristies, aci sys1em relief valve character-istics. The design pressures and data for the respective system compo-nents are listed in Tables 4-2 through 4-6.

4.1.2.2 Design Te'::cerature The design temperature for each co=ponent is selected above the =aximum anticipated coolant te=perature in that co=ponent under cll noz=al and transient load conditions. The design and operning te=peratures of the respective system components are listed in Tables 4-2 through 4-6.

4.1.2 3 Reaction Loads All components in the reactor coolant system are supported and inter-connected so that piping reaction forces result in combined =echanical and thermal stresses in equipment nozzles and structural valls within established code limits. Equipment and pipe supports are designed to absorb piping rupture reaction loads for elimination of secondarf acci-dent effects such as pipe motion and equipment foundation shifting.

4.1.2.4 Seismic Loads Reactor coolant system components are designated as Class I equip =ent, and are designed to maintain their functional integrity during earthquake.

The basic design guide for the seismic analysis is the AEC publication TID-7024, " Nuclear Reactors and Earthquake". Structures and equipment -

vill be designed in accordsnee with Appendix 5A.

4.1.2 5 Cyclic Ioads All components in the reactor coolant system are designed to withstand the effects of cyclic loads due to reactor sys*em temperature and pressure changes. These cyclic loads are introduced by nor=al unit load transients, reactor trip, and startup and shutdown operation. Design cycles are shown in Table k-7 During unit startup and shutdown, the rates of te=perature and pressure changes are 11=ited.

4.1.2.6 Water Chemistrf The water chemistry is selected to provide the necessar/ boron content for l

reactivity control and to minimize corrosion of reactor coolant system surfaces. The reactor coolant chemistry is discussed in further detail in 9 2.

4.1 3 w w au OPERATING CONDITIONS I Throughout the load range from 15 to 100 per cent power, the reactor cool- l ant system is operated at a constant average te=perature. Reactor coolant I system pressure is controlled to provide sufficient overpressure to =ain-tain adequate core subecoling.

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!O The minimum operating pressure is established from core thermal analysis.

This analysis is base.1 upon the maximum expected inlet and outlet temper-

! atures, the maximum reactor power, the minimum DNER required (including

instrumentation errors and the reactor control system deadband), and a core flow distribution factor. The maximum operating pressure is estab-lished on the basis of ASME Code relief valve characteristics and the cargine required for normal pressure variations in the system. Pressure control between the preset maximum and minimum limits is obtained di-rectly by pressurizer spray action to suppress high pressure and pres-surizer heater action to compensate for 1o1. pressure. Normal operational lifetime transient cycles are discussed in detail in 4.1.4.

4.1.4 SERVICK LIFE The service life of reactor coolant system pressure components depends upon the end-of-life material radiation damage, nuclear unit operational thermal cycles, quality manufacturing standards, environmental protection, and adherence t.o established operating procedures. In the following dis-cussion each of these life-dependent factors will be discussel with re-gard to the affected components.

4.1.4.1 Material Radiation Damage The reactor vessel is the only reactor coolant system component exposed to a significant level of neutron irrsdiation and is therefore the only

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' component subject to material radiation damage. To assess the potential radiation damage at the end-of-reactor service life, the m M mum exposure from fast neutrons (E > 1.0 Mev) has been computed to be 3 0 x 1019 n/cm2 over a h0 year life with an 80 per cent load factor. Reactor vessel ir-radiation exposure calculations are described in 3 2.2.1 7 For this neutron exposure, the predicted Nil-Ductility Transition T er-ature (NOIT) shift is 250 F based on the curve shown in F16 ure 4-4. 1 Based on su initial NUIT of 10 F, this shift would result in a predicted NUIT of 260 F.

The " Trend Curve for 550 F Data", as shown in Figure k-4, reprecents ir-radiated material test results and was compiled from the reference docu-ments listed in Table 4-11.

To evaluate the NDTT shift of velds, heat-affected zones, and base mate-rial for the material used in the vessel, test coupons of these three material types have been included in the reactor vessel surveillance program as described in 4.4 3 k.l.4.2 Nuclear Unit Operational Themal Cycles To establish the service life of the reactor coolant syr.com components as required by the ASME III for Class "A" vessels, the nuclear unit operating conditions that involve the cyclic application of loads and thermal con-ditions have been established for the h0-year desi6n life.'

O The number of themal and loading cycles to be used fcr desi n6 purposes are listed in Table 4-7, " Design Cycles". The estimated actual cycles W C.# ,, 0003 323

based on a review of existing nuclear stations operations are also pro-vided in Table 4-7 rable k-9 lists those components designed to ASME III - Class "A". The effect of inditridual transients and the sum of these transients are evaluated to detemine the fatigue usage factor during the detail design and stress analysis effort. As specified in ASME III Para-graph 415 2 (d)(6), the cumulative fatigue usage factor v1H be less than 1.0 for the design cycles listed in Table 4-7 The transient cycles listed in Table 4-7 are conservative and complete in that they include a n significant modes of no mal and emergency operation.

The estimated frequency bases for the design transient cycles are listed in Table 4-8.

Aheatupandcooldownrateof100F/hourvillbeusedintheanalysisof Transients 1 and 2 in Table 4-7 The miscellaneous transients (Item 8) listed in Table 4-7 include the initial hydrotests, plus an allowance for future hydrotests in the event that reactor coolant system modifications or repairs =ay be required.

Subsequent to a r.omal refueling operation only the reactor vessel clo-sure seals are hydrotested for pressure integrity; therefore, reactor coolant system hydrotests before startup are not included.

4.1.4 3 Operating Procedures

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The reactor coolant system pressure vessel components are designed using the transition temperature =ethod of mini =1:ing the possibility of brittle fracture of the vessel =aterials. The various combinations of stresses /

are evaluated and e= ployed to detemine the system operating procedures.

l The basic determination of vessel operation from cold startup and shut-down to full pressure and temperature operation is perfor=ed in accorjance with a " Fracture Analysis Diagram" as published by Fellini and Pu ak.g2)

At temperatures below the nil-ductility transition temperature (NDPI) and the design transition te=perature (DIT), which is equal to NDIT + 60 F, the pressure vessels vin be operated so that the stress levels vin be I

restricted to a value that vill prevent brittle failure. These levels are

a. Below the temperature of DPr minus 200 F, a maximum stress of 10 per cent yield strength.

b. From the temperature of DPI minus 200 F to DPI, a maxiwm stress which vill increase from 10 to 20 per cent yield strength,

c. At the temperature of DPI, a =aximum stress of 20 per cent yield strength.

If the nominal stresses are held within the referenced stress limits (a through e above), brittle fracture vi u not occur. This state = is based on data reported by Robertson(3) and Kihsra and Masubichi in published literature. It can be shown that stress limits can be con-trolled by imposing operating procedures that control pressure and b > N. u_u 0003 324

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O temperature during heatup and cooldown.(3) This procedure vill insure ,

that the nominal stress levels do not exceed those specified in a through c above.

4.1.k.4 Quality Manufacture Material selection is discussed in detail in 4.2 5 After receipt of the material a program of qualification of all velding and heat treating processes that could affect mechanical or metallurgical properties of the material during fabrication is undertaken. This pro-gram vill establish the properties of the material, as received, and certify that the mechanical properties of the materials in the finished vessed are consistent with those used in the design analysis. This pro-gram consists of

a. Weld qualification test plates using production procedures and subjecting test plates to the heat treat =ents to be used in fabricating the vessels.

b. Subjecting qualification test plates to all nondestructive tests to be employed in production, such as x-ray, dye-penetrant, magnetic particle, and ultrasonic. Acceptance standards are the same as used for production.

( c. Subjecting qualification test plates to destm etive tests to establish (1) Tensile strength.

(2) Ductility.

(3) Resistance to brittle fracture of the veld metal, base metal, and heat-affected zone (HAZ) =etal.

After completion of the qualification test program, production velding and inspection procedures are prepared.

All plate or other materials are permaner.tly identified, and the identity is maintained throu6hout manufacture so that each piece can be located in the finished vessels. In-pmcess and final dimensional inspections are made to insure that parts and assemblies meet the draving requirements, and an "as-built" record of these dLiensions is kept for future reference.

All velders are qualified or requalified as necessary in accordance with The Babcock & Wilcox Company and ASME IX requirements. Each lot of veld-ing electrodes and fluxes is tested and qualified before release to in-sure that required mechanical properties and as-deposited chemical prop-erties can be met. Electrodes are identified and issued only on an ap-proved request to insure that the correct materials are used in each veld, n All velding electrodes and fluxes are maintained dry and free from contami-Q nation before use. Records are maintained and reviewed by veldins engineers

, o* insure that appmved procedures and materials are being used. Records

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are =aintained for each veld joint and include the velder's na=e, essen-tial veld para =eters, and electrode heat or lot nu=ber.

The several types of nondestructive tests perfor=ed during vessel fabrica-tion are as follows:

a. Radiography, including x-ray, high voltage linear accelerator, or ndicactive sources, vill be used as applicable to detemine the acceptability of pressure integrity velds and other velds as specificationa require.

b. Ultrasonics is used to examine all pressure-integrity rav mate-rial, the bond between corrosion-resistant cladding to base =a-terial, and pressure-containing velds.

c. Magnetic Particle Examination is used to detect surface or near surface defects in machined veld grooves prior to veldin6, com-pleted veld surfaces, and the ec=plete external surface of the w ssels including veld sea =s after final heat treat =ent.

d. Liquid Penetrant is used to detect surface defects in the veld deposit cladding, non=agnetic =aterials, and closure studs.

The co=pleted reactor vessel assembly vill be shipped as a unit from the fabrication shop to the station site. The completed reactor closure head will be shipped in like =anner.

4.1 5 CODES AND CLASSIFICATIONS All pressure-containin6 co. ponents of the reactor coolant system are de-signed, fabricated, inspected, and tested to applicable codes as listed in Table 4-9 4.2 SYSTEM DESCRIPTION AND OPERATION 4.2.1 GENERAL DESCRIPTION The reactor coolant system consists of the reactor vessel, two vertical once-through steam generators, four shaft-sealed coolant circulating i

pu=ps, an electrically heated pressuri er, and interconnectin6 piping.

l The system is arranged as two heat transport loops, each with two circu-

! lating pt ps and one steam generator. Reactor system design data are listed in Tables 4-2 through 4-6, and a system sche =atic dia6 ram is shown in F Qure 4-1. Elevation and plan views of the arrange =ent of the major co.penents are shown in Figures 4-2 and 4-3 4.2.2 MAJOR CCMPONENTS 4.2.2.1 Reactor Vessel The reactor vessel consists of a cylindrical shell, a cylindrical support skirt, a spherically dished bottom head, and a rin6 flange to which a re-movable reactor closure head is bolted. The reactor closure head is a spherically dished head welded to a ring flange.

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rs The vessel has six =ajor no::les for reactor coolant flow, 69 control rod drive no::les =ounted on the reactor closure head, and two core flooding no::les--all located above the core. The vessel closure seal is for=ed by two concentric 0-rings with provisions between them for leak detection.

The reactor vessel, no::le design, and seals incorporate the extensive design and fabrication experience accu =ulated by MW. lifty-two in- 1 core instru=entation no::les are located on the lover head.

The reactor closure head and the reactor vessel flange are joined by sixty 6-1/2 in. diameter studs. Two =etallic 0-rings seal the reactor vessel when the reactor closure head is bolted in place. Pressure taps are pro-vided in the annulus between the two 0-rings to =enitor leakage and to hydrotest the vessel closure seal after refueling.

The vessel is insulated with =etallic reflective-type insulation. Insu- 1 lation panels are provided for the reactor closure head.

The reactor vessel internals are designed to direct the coolant flow, sup-port the reactor core, and guide the control rods in the withdrawn posi-tion. The reactor vessel cantains the core sunport asse=bly, upper ple-num asse=bly, fuel assemblies, centrol rod assemblies, surveillance speci-mens, and incore instru=entation.

The reactor vessel shell =aterial is protected a611nst fast neutron flux and gn-ma heating effects by a series of water annuli and the the=al

'p shield located between the core and vessel vall. This protection is fur-

, V ther described in 3 2.4.1.2, 4.1.4, and 4 3 1.

Stop blocks velded to the renetor vessel inside vall limit reactor inter-nals and core vertical drop to 1/2 in. or less and prevent rotation about the vertical axis in the unlikely event of a major internals component failure.

Surveillance specimens made'from reactor steel are located between the reactor vessel vall and the ther=al shield. These speci= ens vill be ex-amined at selected intervals to evaluate reactor vessel =aterial ?mIT changes as described in 4.k.3 The reactor vessel general arrangement is shown in Figure 4-5, and the general arrangement of the reactor vessel and internals is shown in Fig-ures 3-45 and 3-46. Reactor vessel design data are listed in Table L-2. l1 4.2.2.2 Pressurizer The general arrangement of the reactor coolant system pressurizer is shown in Figure 4-6, and the design characteristics are tabulated in Table 4-3 The electrically heated pressuri er establishes and maintains the reactor coolant pressure within prescribed limits and provides a surge chamber and a water reserve to acco=modate reactor coolant volu=e changes during operation.

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outlet piping by the surge piping. The pressurizer vessel is protected h() (' , ,

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frem ther=al effects by a ther=al sleeve on the surge line and by a dis-tribution baffle located above the surge pipe entrance to the vessel.

Relief valves are mounted on the top of the pressuri:er and function to

! relieve say system overpressure. Each valve has one-half the required relieving capacity. The capacity of these valves is discussed in 4 3.4.

The relief valves discharge to a reactor coolant drain tank located with-in the reactor building. The drain tank has a stored water supply to condense the steam. A relief valve protects the tank against overpres-sure should a pressurizer valve fail to rescat.

The pressurizer contains Jsplaceable electric heaters in its lower sec-tion and a water spray ne.rt.le in its upper section to =aintain the steam and vater at the saturation te=perature corresponding to the desired re-actor coolant system pressure. During outsurges, as the pressure in the reactor decreases, seme of the water in the pressurizer flashes to steam t to maintain pressure. F.lectric heaters are actuated to restore the nor-

, mal operating pressure. During insurges, as pressure in the reactor sys-l tem increases, steam is condensed by a vater spray from the reactor inlet I lines, thus reducing pressure. Spray flow and heaters are controlled by I the pressurizer pressure controller.

l Instrumentation for the pressurizer is discussed in 7 3 2.

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4.2.2 3 Steem Generator

"'he general arrangement of the steem generators is shown in Figure L-7, O j and design data are tabulated in Table h-h.

The steam generator is a vertical, strsight-tube-and-shell heat exchanger and produces superheated steam at constant pressure over the power range.

Reactor coolant flows downward through the tubes, and steam is generated on the shell side. The high pressure parts of the und t are the hemisphe-rical heads, the tubesheets, and the straight Inconel *) tubes between the tubesheets. Tube supports hold the tubes in a unifoz : pattern along their length.

The shell, the cutside of the tubes, and the tubesheets form the boundaries of the steam-producing section of the vessel. '41 thin the shell, the tube bundle is surrounded by a shroud, which is divided into two sections.

The upper part of the annulus between the shell and baffle is the super-heater outlet, and the lover part is the feedvater inlet-heating zone.

Vents, drains, instrumentation no::les, and inspection openings are pro-

vided on the shell side of the unit. The reactor coolant side has in-l strumentation connections en the top and bottom heads, =anvays on both l heads, and a drain no

:le for the bcttom head. Venting of the reactor Inconel is a trade na=e cf an alloy canufactured by the International i Nickel Co.~pany. It also hrus substantial ecm=on usage as a generic de-scription cf a Ni-Fe-Cr alloy conforming to ASTM Specification SB-163 It is in the la.ter context that reference is =ade here.

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O coolant side of the unit is accomplished by a vent connection on the re-actor coolant inlet pipe to each unit. The unit is supported by a skirt attached to the bottom head.

Reactor coolant water enters the steam generator at the upper plenum, flows down the Inconel tubes while transferring heat to the secondary shell-side fluid, and exits through the lower plenum. Figure 4-8 shows the flow paths and steam generator heating regions.

Four her.t transfer regions exist in the steam generator as feedvater is converted to superheated steam. Starting with the feedvater inlet these are

a. Feedvater Heating Feedvater is heated to saturation temperature by direct contact heat exchangt The feedvater entering the unit is sprayed into a feed heating annulus (downcomer) formed by the shell and the baffle around the tube bundle. The steam that heats the feed-vater to saturation is drawn into the downcomer by condensing action of the relatively cold feedvater.

The saturated water in the downcomer forms a static head to balance the static head in the nucleate boiling section. This provides the head to overcome pressure drop in the circuit fozmed by the do n .ucr, tne boiling sections, and the bypass steam flow to the 'mivater heating region. With lov (less than 1 ft/sec) satuitted water velocities entering the gener-ating section, the secondary side pressure drops in the boil-ing section are negligible. The majority of the pressure drop is due to the static head of the mixture. Consequently, the downcomer level of water balances the =ean density of the two-phase boiling mixture in the nucleate boiling region.

b. Nucleate Boiling The saturated water enters the tube bundle, and the stear-water mixture flows upward on the outside of the Inconel tubes counter-current to the reactor coolant flow. The vapor content of the mixture increasec 'almost uniformly until DNB, i.e. , departure of nucleate boiling, is reached, and then film boiling and super-heating occurs. The quality at which transition from nucleate j ho111ng to film boiling occurs is a function of pressure, heat flux, and mass velocity,

c. Film Boiling DI7 saturated steam is produced in the film boiling region at the upper end of the tuce bundle.

d. Superheated Steam .

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Saturated steam is raised to final temperature in the super- l heater region.

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Shown on Figure 4-9 is a plot of heating surface and downcocer level versus load. As shown, the downcomer vater level is proportional to stes= flow from 15 - 100 per cent load. A constant minimum level is held below 15 per cent load. The amount of surface (or length) of the nuc1cate boiling section and the film boiling section is proportional to load. The sur-face available for superheating varies inversely with load, i.e., as load decreases the superheat section gains from the nucleate and film boiling regions.

Mass inventcij in tne steam generator increases with load ac the length of the heat transfer regions varies.

The si@ le concept with ideal counterflow conditions results in highly stable flow characteristics on botn tne reactor coalant and secondary sid<s. The hot reactor coolant fluid is cooled unifomly as it fiova downward. The secondary side = ass flow is lov, and the =ajority of the pressure drop is due to the static effect of the mixture. The boiling in the stean generator is somewhat similar to "pcol boiling", except that there is motion upward that permits so=e parallel flow of water and steam.

A plot of reactor coolant and steam te=peratures versus power is shown in Figure 7-5 As shown, both steam pressure and average reactor coolant te=perature are held constant over the load range from 15 to 100 per cent full power. Constant steam pressure is obtained by a variable two-phase boiling length (see Figure 4-8) and by the regulation of feed flow to ob-tain proper steam generator secondary =ess inventot/. In addition to average reactor coolant te=perature, reactor coolant flow is also held '

constant. The difference between reactor coolant inlet and outlet tem-peratures increases pretorticnately as load is increased. Saturation pressure and temperat ae are constant, resulting in a variable superheater outlet te=perature.

Figure 4-10, a plot of te=perature versus tube length, shows the te=pera-ture differences between snell and tube throughout the steam generator at fu'J load. The excellent heat transfer coefficients per=it the use ot' a secur.dary operating pressure and te=perature sufficiently close to the rer.ctor coolant average temperature so that a straight-tube design can be used.

The shell te=perature is controlled by the use of direct contact steam that heats the feedvater to saturation, and the shell is bathed with sat-( urated water from teedvater inlet to the lower tubesheet.

1 In the supernester section, the tube vall te=perature approaches the re-actor coolant fluid temperature since the steam film heat transfer coef-ficient is considerably lover than the reactor coolant heat transfer coef-l ficient. By baffle arrange =ent in the superheater section, the shell see-l tion is bathed with superheated steam above the steam outlet noccle, fur-l ther reducin6 te=perature differentials be:veen tubes and shell.

The steam generator design and stress analysir, vill be perfomed in ac-cordance with the require =ents of the A m III as described in 4 3 1.1. $

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4.2.2.h Reactor Coolant Pumps The general arrangement of a reactor coolant pump is shot n in Figure 4-11, and the pump design data are tabulated in Table 4-5 The reactor coolant pumps are vertical, single-speed, shaft-sealed units having bottom suc-tion and horizontal discharge. Each pu=p has a separate, single-speed, top-mounted motor, which is connected to the pump by a shaft coupling.

Shaft sealing is accomplished in the upper part of the pump housing using a throttle bushing, a seal cha=ber, a mechanical seal, and a drain chamber in series. Seal water is injected ahead of the throttle bushing at a pressure approximately 50 psi above reactor system pressure. Part of the seal flow passes into the pump volute through the radial pump bearing.

The remainder flows out along the throttle bushing, where its pressure is reduced, to the seal chamber and is returned to the seal water supply system. The outboard mechanical seal normally operates at a pressure and temperature of approximately 50 psig and 125 F. However, it is designed for full reactor coolant system pressure and, if seal chamber cooling were maintained, would continue to operate satisfactorily without seal water injection for several weeks. The outboard drain chamber would fur-ther prevent leakage to the reactor building if deterioration of the me-chanical seal performance should occur.

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 pump motor. The I

thrust bearing is designed so that reverse rotation of the shaft will not lead to pump or motor damage. Lube oil cooling is accomplished by cooling coils in the motor oil reservoir. 011 pressure required for bearing lub-rication is maintained by internal pumping provisions in the motor, or by and external system if required for " hydraulic-jacking" of the ber. ring surfaces for startup.

An antirotation device will be furnished with each pump motor to prohibit reverse rotation of the pu=p.

Factory thrust, vibration, and seal performance tests will be made in a closed loop on the first pump at rated speed with the pump end at rated temperature and pressure. Sufficient testing will be done on subsequent units to substantiate that they conform to the initial test pump charac-teristics k.2.2.5 Reactor Coolant Piping The general arrargament of the reactor coolant system piping is shown in Figures k-2 and L-3 Piping design data are presented in Table b-6. In addition to the pressuriser surge piping connection, the piping is equipped with welded connections for pressure taps, temperature elements, vents, drains, decay heat removal, and emergency core cooling high pres-sure injection water. Thermal sleeves are provided in the pressurizer

. urge piping and the emergency high pressure injection pipe connections.

o) 9 I

b\ \

' ' s u-u 000] 031 l

k.2 3 FRESsUPI-PILIE7ING DEVICES The reactor coolant system is protected against overpressure by control O

and protective circuits such as the high pressure trip and code relief valves located on the top head of the pressurizer. The relief valves discharge into the reactor coolant drain tank which condenses and collects the effluent. The schematic arrangement of the relief devices is shown in Figure L-1. since an sources of heat in the system, i.e., core, pres-eurizer heaters, and reactor coolant pu=ps, are interconnected by the re-actor coolant piping with no intervening isolation valves, au relief pr'o-tection can conveniently be located on the pressurizer.

h.2.h ENVIRONMENTAL PRCfIECTION The reactor coolant system is surrounded by concrete shield vans. These vans provide shielding to permit access into the reactor build!ng for inspection and maintenance of miscellaneous rotating equip =ent during rated power operation and for periodic calibration of the incere =onitor-ing system. These shielding valls act as missile protection for the re-actor building liner plate.

Lateral bracing vill be provided near the steam generator upper tubesheet elevation to resist lateral loads, including those resulting from seis=ic forces, pipe rupture, themal expansion, etc. Additional bracing is pro-vided at a lcwer elevation to restrain the 36 in. ID vertical pipe les from whipping.

Barriers over the reactor coolant system are also provided for shielding 1 arai missile damage prctection. ,

4.2 5 MATERIAIS CF CONSTRUCTICN Each of the materials used in the reactor coolant system has been selected for the expected environ =ent and service conditions. The major ec=ponent

=aterials are listed in Table L-10.

All reactor coolant system =aterials exposed to the coolant are corrosion-resistant =aterials consisting of 304 or 316 ss, veld deposit 304 ss clad-ding, Inconel (Ni-Cr-Fe), and 17-k PH (H1100). These materials were chosen for specific purposes at various locations within the system because of their superior co=patibility with the reactor coolant.

Periodic analyses of the coolant chemical composition vill be perfor=ed to

=enitor the adherence of the system to the reactor coolant water quality listed in Table 9-h. Maintenance of the water quality to minimize corro-sion is perfor=ed by the che=ical addition and sa=pling system which is described in detail in 9 2.

The feedvater quality entering the steam generator vill be held within the limits listed in Table 9-3 to prevent deposits and corrosion inside the steam generators. This required feedvater quality has been successfully used in co= parable once-through, nonnuclear stea= gensrators. The phe-no=ena of stress-corrosion cracking and corrosion fatigue are not gener-ally encountered unless a co=bination of elements in varying degrees is t.o.

4-12 (Revised 7-21-c7) 01100.332

. s

(

present. The necessar/ conditions are a susceptible alloy, an aggressive environment, a stress, and time.

A n external insulation of reactor coolant system components vill be com-patible with the component materials. The reactor vessel is insulated with metanic reflective insulation on the cylindrical shen exterior.

The closure flanges and the top and bottom heads in the area of corrosion-resistant penetrations wi n be insulated with low-halide-content insulating

=aterial. An other external corrosion-resistant surfaces in the reactor coolant system vin be insulated with lov or halide-free insulating cate-rial as required.

The reactor vessel plate material opposite the core is purchased to a specified Charpy V-notch test result of 30 ft-lb or greater at a corre-sponding nil-ductility transition temperature (NUIT) of 10 F or less, and the material win be tested to verify conformity to specified re-quirements and to determine the actual NITI'T value. In addition, this plate vin be 100 per cent volumetricany inspected by ultrasonic test ,

using both normal and shear wave. I The reactor vessel material is heat-treated specifically to obtain Scod notch-ductility which win insure a lov ICIT and thereby give assurance that the finished vessel can be initially hydrostatically tested and op-ersted at room temperature without restrictions. The stress limits es-tablished for the reactor vessel are dependent upon the temperature at which the stresses are applied. As a result of fast neutron absorption in the region of the core, the material ductility vill change. The effect is an increase in the NDIT. The predicted end-of-life NIFIT value of the reactor vessel opposite the core is 260 F or less. The predicted neutron exposure and IMIT shift are discussed in 4.1.4.

The unirradiated or initial NDIT of pressure vessel base plate material is presently measured by two methods: the drop veight test given in ASTM E208, and the Charpy V-notch impact test (Type A) given in ASTM E23 The NDIT is defined in ASTM E208 as "the temperature at which a specimen is broken in a series of tests in which duplicate no break performance occurs at a 10 F higher temperature". Using the Charpy V-notch test, the NDIT in defined as the temperature at which the energy required to break the specimen is a certain " fixed" value. For SA 302B steel the ASME In Table N-332 specifia an energy value of 30 ft-lb. This value is based on a correlation with the drop veight test and win be referred to as the "30 ft-lb fix". A curve of the temperature versus energy absorbed in breaking the specimen is plotted. To obtain this curve, 15 tests are perfor=ed which include three tests at five different temperatures. The intersection of the energy versus temperature curve with the 30 ft-lb ordinate is designated as the NUIT.

The available data indicate differences as great as 40 degrees between curves plotted through the mini =um and average values respectively. The determination of NITIT from the average curve is considered represet.tative of the material and is consistent with procedures specified in ASTM E23

( In assessing the NDIT shift due to irradiation, the translation of the average curves is used.

y'. . v.

h-13 0000 353

I The =aterial for these tests vill be treated by the =ethods outlined in ASME III Paragraph N-313 The test coupons vill be taken at a distance of T/4 (1/4 of the plate thickness) from the quenched surfaces and at a distance of T frcm the quenched edges. These tests are perfor=ed by the

=aterial supplier to certify the =aterial as delivered to MW. The exact test coupon locations are reviewed and approved by MW to insure compli-ance with the applicable AS!E Code and specifications. In accordance with AS!E III Paragraphs N-712 and N-713, MW perfor=s Charpy V-notch i= pact tests on heat-affected cone (HAZ), base =etal, and veld =etal on all pressure vessel test plates.

Differences of 20 to 40 F in NDIT have been observed between T/4 and the surface in heavy plates. The T/4 location for Charpy V-notch impact specimens is conservative since the NDIT of the surface =aterial is lower than that of the internal caterial.

The reactor vessel design includes surveillance speci= ens which vill per-mit an evaluation of the neutron exposure-induced shift on the =aterial sil-ductility transition te=perature properties.

The re=aining material in the reactor vessel and the other reactor cool-ant system components are purchased to the appropriate design code re-quire =ents and specific co=ponent fun,ction.

The =aterial irradiation surveillance program is described in k.4 3 4.2.6 MAXIMUM HEATING AND CCOLING RATES The corral reactor coolant system operating cycles are given in Tables 4-7 and 4-8 and described in 4.1.4. The normal system heating and cool-ingrateis100F/hr. The exact final rates are determined during the detail design and stress analysis of the vessel.

The fastest cooldown rates resulting from the break of a =ain steam line are discussed in 14.1.2 9 l

l 4.2 7 LEAK DEIECTION l

l To minimize leakage from the reactor coolant system all components are

interconnected by en e.11-velded piping system. Some of the co=ponents I have access openings of a flanged-gasketed design. The largest of these is the reactor closure head, which has a double =etal 0-ring seal with provisions for monitoring for leakage between the 0-rings. Other open-ings a.d appu m nances to the reactor coolant system that are possible sources of leakage are tabulated in detail along with the =ax1=um ex-pected rates of leckage in Se6 tion 11.

With regard to the reactor vessel, the probability of a leak occurring is considered to te remote on the basia of reactor vessel design, fabrication, test, inspection, and operation at terperatures above the =aterial NDIT as described in 4 3 1. Reactor closvro hM leakage vill be cero from the annulus between the =etallio 0-rius seals during vessel steady-state and virtually all transient operating conditions. Only in the event of a rapid transient operation, such as an e=ergency cooldown, would there be 4-14

some leakage past the innermost 0-ring seal. A stress analysis on a sim-ilarvesseldesignindicatesthisleakratevouldbeapproximately10cc/

min throu6h the seal monitoring taps to a drain, and no leakage vin occur past the outer 0-ring seal. The exact nature of this transient condition and the resulting sman lesk rate vin be dete=ined by a detailed stress analysis.

In the unlikely event that an extensive leak should occur from the system into the reactor buildins during reactor operation, the leakage vin be detected by one or more of the followin6 methods:

a. Instrumentation in the control room vi n indicate the addition rate of makeup water required to maintain normal vater level in the pressurizer and in the makeup tank. Deviation from nonnal makeup and letdown to the resctor coolant system will provide l1 an indication of the magnitudo of the leak.

b. Control room instrumentation vin indicate additional reactor building atmosphere particulate or radioactive gas activity.

c. Control room instrumentation vill indicate the existence of a change in the water level in the reactor buildin6 sump.

If any one of the methods above indicates an excessive reactor coolant leakage rate during operation, the reactor will be taken to a cold sh it-down, and the cause of the problem vill be determined.

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43 SYSTEM DESIGN EVALUATION 431 SAFlTl FACTORS The reactor coolant system is designed, fabricated, and erected in accor-dance with proven and recognized design ecdes and quality standards appli-cable for the specific co=ponent function or classification. Ihese e m-ponents are designed for a pressure of 2,500 psig at a nominal te=pera-ture of 650 F. The corresponding nominal operating pressure of 2,185 psig allove an adequate =argin fox' non::a1 lead changes and operating transients. The reactor system cecponents are designed to meet the codes listed in Table 4-9 Aside from the safety factors introduced by code require =ents and quality control programs, as described in the following paragraphs, the reactor coolant system functional safety factors are discussed in Sections 3 and 14.

4 3 1.1 Pressure vessel Safety The safety of the nuclear reactor vessel and all other reactor coolant system pressure vessels is dependent upon four =ajor factors: (1) de- 1 sign and stress analysis, (2) quality centrol, (3) prcper operatica, and (L) relief valves. The special care and detail used in i=ple=enting these factors in pressure vessel manufacture are briefly described as follows:

k.3 7. 1.1 Design and Stress Analysis Thest. 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 frem industry and govern =ent who are skilled in the design and fabrication of pressure vessels. It is a comprehensive code based on the =ost appli-cable stress theory. It requires a stress analysis of the entire vessel under both steady-state and transient operations. The result is a com-plete evaluation of both pri= arf and secondarf stresses, and the fatigue life of the entire vessel. This is a contrast with previous codes which basically established a vessel thickness during steady-state operations only. In establishing the fatigue life of these pressure vessels, using the design cycles frem Table 4-7, the fatigue evaluation curves of ASME III are e= ployed.

Since ASME III requires a complete stress analysis, the designer =ust have at his disposal the necessarf analytical tools to acco=plish this. These tools are the solutions to the basic =athe=atical theory of elasticity equations. In recent years the capability and use of ec=puters have played a major part in refining these analytical solutions. The Babcock

& Wilcox Company has confir=ed the theory of plates and shells by =casur- ,

ing strains and rotations on the large flanges of actual precsure vessels and finding them to be in agree =ent with those predicted by the theerf.

B&W has also conducted laboratorf deflection studies of thick shell and ring combinations to define the accuracy of the theory, and is using ecm-puter programs developed on the basis of this test data.

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.t' 4-16 (Fevised 7-21-67) .

0000 336 e w.

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The analytical procedure considers all process operation conditions. A detailed design and analysis of every part of the vessel is prepared as fellows:

a. The vessel size and configuration are set to meet the process requirements, the thickness requirements due to pressure and othsr structural dead and live loads, and the special fillet contour and transition taper requirements at nozzles, etc., re-quired by ASME III.

b. The vessel pressure and temperature design transients given in Table 4-7 are employed in the detemination of the pressure loading and temperature gradient and their variations with time throughout the vessel. The resulting combinations of pressure loading and thermal stresses are calculated. Computer programs are used in this development.

c. The s+resses through the vessel are evaluated using as criteria the allowable stresses per ASME III. This code gives safe stress level limits for all the types of applied stress. These are membrane stress (to insure adequate tensile strength of the vessel), membrane plus primary bending stress (to insure a dis-tortion-free vessel), secondary stress (to insure a vessel that O

U vill not progressively deform under cyclic loading), and peak stresas (to insure a vessel of maximum fatigue life).

A design report is prepared and submitted to the jurisdictional authort-ties an.1 regulatory agencies, i.e., state, insurance, etc. This report defines in sufficient detail the design basis, loading conditions, etc.,

and will summarize the conclusions to permit independer+ checking by interested parties.

4 3 1.1.2 quality control In-process and final dimensional inspections are =ade to insure that parts ard assemblies meet the drawing requirements, and an "as-built" record of these dimensions is kept for reference. A temperature-controlled gage-room is maintained to keep all measuring equipment in proper calibration, ard personnel supervisir.g this work are trained in femal programs spon-sored by gage equipment manufacturers.

The practice of applied radiography is being continually i= proved to en-hance flav detecticn. Present procedures are

a. All velds are properly prepared by chipping and grinding valleys between stringer beads so that radicgraphs can be properly in-terpreted.

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b.

O All radiographs are reviewed by two people knowledgeable and skilled in their interpretation.

c. An 0.080 in. lead filter is used at the fi1= to absorb " broad-bea= scatter" when usir4 nigh voltage equip =ent (above 1 Mev).

d. Fine grain er extra fine grain fil is used for all exposures.

e. Densities of radiographs are centrciled by densitc=eters.

f. Double fil: technique is used on all gn==a-ray exposures as well as high voltage exposures.

g. 711=s are processed through an autc=atic processor which has a controlled replenish =ent, te=perature, arai process cycle, all centributing te better quality,

b. Energies are controlled so as to be in the opti=u= range.

Ultrasonics, one of the =ost useful of inspection tools, is being used as follcvs:

a. In addition to radiography, pressure-containing velds are in-spected by ultrasonics.

b. In order to detect la=inations which are nor= ally parallel to the surface, plates are also inspected by a shear vave,

c. The bond between cladding anii base =aterial is inspected by ultrasonics.

d. All plate is 100 per cent volu=etrically inspected by ultrasenics using both nom al and shear wave.

e. personnel conducting ultrasonic inspections are given extensive training.

De =agnetic particle examination is used to aid in detecting surface and near surface defects and is e= ployed on both parts and finished vessels as follows:

a. Welds are inspected with the =agnetic particle =ethod after re-

=ovs1 of backup strips.

b. Weld preparations are inspected by the =agnetic particle =ethod.

c. De external surface of the entire vessel, including veld see=s, is inspected after all heat treat =ent.

d. Persennel using this =ethod are trained by S&W and by the equip-ment =anufacturers who offer foc al training progra s.

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O The liquid penetrant examination helps to detect surface defects and is particularly adaptable to the non=agnetic materials such as stainless l

steel. It is presently being used as follevs: 1

a. Inspection of veld-deposited cladding.

b. Inspection of reactor vessel studs.

c. Personnel using this method of examinatica are trained by B&W and by the equip =ent =anufacturers.

The pri=ary purpose of these qualfty control procedures and methods is to locate, define, and determine the size of material defects to allow an evaluation of defect acceptance, rejection, or repair.

The size of defect that can conceivably contribute to the rupture of a vessel depends not only on the size effect but on the orientation of the defect, the magnitude of the stress field, and temperature. Dese =ajor parameters have been correlated by Pellini and Pucak(2) who have prepared.

a " Fracture Analysis Diagram" which is the basis of vessel operation frcm cold startup and shutdevn to full pressure and temperature operation.

The diagram predicts that, for a given level of stress, larger flav sizes vill be required for fracture initiation above the Im te=perature. For example, at stresses in the order of 3/4 yield strength, a flaw in the order of 8 to 10 in. =ay be sufficient to initiate fracture at temperatures below the IM temperature. However, at I m *30 F, a flav of 1-1/2 times this size may be required for initiation of fracture. While at a tempera-ture of Im + 60 F, brittle fracture is not possible under elastic stresses because brittle fracture propa6ation does not take place at this temperature. Fractures above this temperature are of the predcminantly ductile type and are dependent upon the = ember net section area and sec-tion =edulus as they establish the applied stress.

Stud forgings will be inspected for flaws by two ultrasonic inspections.

An axial longitudinal beam inspection vill be perfor=ed. The rejection standard vill be loss-of-back-reflection greater than that frem a 1/2 in.

diameter flat bottcan hole. A radial inspection vill be made using the longitudinal beam technique. B is inspection vill carry the same rejec-tion standards as the axial inspection. In addition to the ultrasonic tests, liquid-penetrant inspection vill be perfor=ed on the finished studs.

The stress analysis of the studs vill include a fatigue evaluation. It is not expected that fatigue evaluation vill yield a significantly high usage factor for the h0-year design life. Therefore, there vill be no ple.nned frequency for stud replacement. If an indication is found when the studs are inspected during refueling, as described below, the stud vill be re-placed.

One-third of the studs vill be visually examined and dye-penetrant in-

.spected at every refueling. Any positive indications found vill be cause O for rejection.

,n:

0000 339

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S' h-19

The reactor closure head is attached to the reacter vessel with sixty 6-1/2 in. dia=eter studs. me stud =aterial is A-540, Grade B23 (ASME III, Case 1335) which has a =ini=u::: yield strength of 130,000 psi. The studs, when tightened for operating conditions, v1H have a tensile stress of approximately 30,000 psi. Thus, at operating conditions (2,185 psig):

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

b. 25 adjacent studs can fail before the re=nieng studs reach yield strength,

c. 26 adjacent studs can fail before the re=aining studs reach the ulti= ate tensile strength.

d. 43 sy==etrically located studs can fail before the re=ainin; e

studs reach yield strength.

l 4.3 1.1 3 Operation As previously =entioned in 4.1.4, pressure vessel service life is depen-dent on adherence to established operating procedures. Pressure vessel safety is also dependent on proper vessel operation. Therefc::e, particu-lar attention is given to fatigue evaluation of the pressure vessels and to the factors that affect fatigue life. The fatigue criteria of ASME l III are the bases of designing for fatigue. They are based en fatigue l tests of pressure vessels sponsored by the AEC and the Pressure Vessel Research Cc==1ttee. The stress limits established for the pressure ves-g,'

sels are dependent upon the te=perature at which the stresses are applied.

As a result of fast neutren absorption in the region of the core, the re-actor vessel =aterial ductility vin change. The effect is an increase in the nil-ductility transition te=perature (UDTT). h e dete::lination of the predicted NDTT shift is described in 4.1.4.1. 21s NDTT shift is l

factored into the plant startup and shutdown procedures so that full cper- _

l ating pressure is not attained until the reacter vessel te=perature is above the design transition te=perature (DTI). Belev the IYrT the total stress in the vessel vall due to both pressure and the associated heatup and cooldown transient is restricted to 5,000 - 10,000 psi, which is be-lov the threshold of concern for safe operation. These stress levels de-fine an operating coolant pressure te=perature path or envelope for a stated heatup or cooldown rate that =ust be followed. Additional infor-l =ation on the deter =ination of the operating procedures is provided in 4.1.4.1, 4.1.4.2, and 4.1.4 3 4 3 1.1.4 Additional Pressure Vessel Safety Factors Additional =ethods and precedures used in pressure vessel design, not previously =enticned in 4 31.1 above but which are considered conserva-tive and provide an additional =argin of safety, are as follevs:

a. Use of a stress concentration factor of 4 cn assu=ed flave in calculating stresses.

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u-20 0000 4 0 -

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b. Use of min 4 mn= specified yield strength of the material instead of the actual values.

c. Neglecting the increase in yield strength resulting from irra-diation effects.

d. The design shift in NDTT as given in 4.1.4.1 is based on maxi-mum predicted flux levels at the reactor vessel inside vall sur-face, vbereas the bulk of the reactor vessel material vill ex- l perience a lesser exposure of radiation and consequently a lover change in NDTT over the life of the vessel.

e. Because the irradiation dosage is higher at the inside surface of the reactor pressure vessel vall, the surveillance specimens will be subjected to a greater degree of irradiation and there-fore to a larger shift in NDTT value than vill be experienced by the vessel. The specimens lead the vessel with respect to irradiation effect.s and impart a degree of conservatism in the evaluation of the capsule specimens. The material irradiation surveillance program is described in 4.4.3

f. Results from the method of neutron flux calculations, as de-scribed in 3 2.2.17, have increased the flux calculations by a factor of 2 in predicting the nyt in the reactor vessel vall.

The conservative asstsptions, uncertaintics, and comparisons of O csiculational codes used in determining this factor are dis-cussed in detail in 3 2.2.1 7 The foregoing discussion presents a detailed description of quality design, fabrication, inspection, and operating precedures used to insure confidence in the integrity of pressure vessels. Experience reported by Reference (5),

l B&W, an.1 the satisfactory experience of MW customers support the conclu-sion that pressure vessel rupture is incredible.

4 3 1.2 Piping Total stresses resulting from' thermal expansion and pressure and mechani-cal and seismic loadings are all considered in the design of the reactor coolant piping. The total stresses that can be expected in the piping are within the marimum code allowables. 'Ihe pressurizer surge line con-

nection and the high pressure injection connecticus are equipped with thermal sleeves to limit stresses from thermal shock to acceptchle values.

All materials and fabrication procedures will meet the requirements of the specified code. All material vill be ultrasonically inspected. All velds will be radiographically" inspected. All interior surfaces of the inter-connecting piping are clad with stainless steel to eliminate corrosion problems and to reduce coolant contamtnation.

4313 Steam Generator i

Because the basic concept of the once-through steam generator vould indi-cate the possible existence of differential thermal expansion-induced str,ypses in either the tubes or shell, the thermal loadings have been

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evaluated using the most severe design transients from Table 4-7 0000 241 I i!

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k-21

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The basic structural premise of the steam generator is that the tubesheeti.

O themselves are' designed to take the full design pressure on either side of the tubesheet with zero pressure on the other side. That is, the cubes are not counted upon for any structural aid er support.

The steam line failure am1v:ed in 14.1.2 9 closely s1=ulates this' design premise in a transient manner. Secondary te=perature variations during the accident are essentiall'y transient skin effects with the controlling temperature for the tubesheets and tubes being that of the reactor cool-ant. Ther=al stresses for this case vin be below ASME allevable values.

Scme tube defor=ation cay occur but will be restrained by the tube sup- t ports.

During normal power operation the tubes are hotter than the she n of the steam generator by 10 to 20 F depending on load. The effect is to put the tubes in a slight compression of 3,000 psi at the 20 F =axL,um tem-perature difference.

During startup and shutdown operations the tubes are hotter than the shell of the steam generator by h0 F. This places the tubes in a ec=pressive stress of 6,000 psi. Thus, the stress levels developed during nor=al startup or shutdown operations cause no adverse effect on the tubes since these stresses are well below the allowable stress of 23,300 psi for this SB-163 material (ASME III, Case 1336). 32ekling of the tubes does not occur since they are supported laterally at 40-in. intervals along their length. To demonstrate the structural adequacy of the steam generator at

. this condition, a laboratory unit was constructed of the same tube size, length, and =aterial as the steam generator, but of seven tubes in number.

It was structurally tested with a thermal difference of shell and tube of 80 F for 2,000 cycles. This severe thermal cycle test was perfor=ed with a tube-to-shell temperature difference tvice as great as the ~ v4 m ex-pected during startup and shutdown (Transients 1 and 2, Table 4-7). De-structive examination of the unit after this test indicated no adverse effects frcm fatigue, stress, buckling, or tube-to-tubesheet joint leak-age.

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k-22 0000 14?*'

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O 432 RELIANCE ON INIZRCONNECTED SYSTEMS The principal heat removal systems which are interconnected with the reac-tor coolant system are the steam and feedvater systems and the decay heat removal system. The reactor coolant system is dependent upon the steam generators, and the steam, feedvater, and cornensate systems for decay heat removal fra nonnal operating conditions to a reactor coolant tem-perature of approximately 250 F. All vital active components in these systems are duplicated for reliability purposes.

i D e engineering flow disgram of the steam and feedvater systems is shown in Figure 10-1. In the event that the condensers are not available to receive the steam generated by decay heat, the water stored in the feed-water syste:n may be pumped into the steam generato2:,s and the resultant steam vented to the atmosphere. A turbine-driven, 5 per cent capacity boiler feed pump vill supply water to the steam generators.

7he decay heat removal system is used to remove decay heat when the ther-mal driving head of the reactor volt.a.t system is no longer adequate to generate steam. This system is completely described in 9 5 Se heat received by this system is ultimately rejected to the nuclear services cooling water system which also contains sufficient redundancy to guaran-tee proper operation. A schematic diagram of the nuclear services cool-ing water system is presented in Figure 9-10.

433 SYSTEM INTEGRITY The integrity of the reactor coolant system is insured by proper materials selection, fabrication quality control, design, and operatien. All ecmpo-nents in the resetor coolant system are fabricated from materials initially having a lov nil-ductility transition temperature (NDTT) to eliminate the possibility of propagating-type failures. Where material properties are subject to change throughout unit lifetime, such as the case with the re-actor vessel, provisions are included for materials surveillance speci-

nens. Rese vill be periodically examined, and any required temperature-Pressure restrictions will be incorporated into reactor operation to in-sure operation above NDTT.

S e reactor coolant system is designed in accordance with ASME pressure vessel aci ASA power piping codes as covered in k.l. Relief ~ valves on the pressurizer are sized to prevent system pressure fr a exceeding the de-sign point by more than 10 per cent.

As a further assurance of system integrity, all components in the system will be hydrotested at 3,125 psig before initial operation.' The largest and most frequently used opening in the reactor coolant system, the reac-tor closure head, contains provisions for separate hydrostatic pressuriza-tien between the 0-ring type gaskets.

434 PRESSURE RELIEF 2 e reactor coolant system is protected against overpressure by safety valves located on the top of the pressurizer.

[j 4 ..

000 343 ^

k-23

The capacity of the pressurizer safety va'.ves is deter =ined frcm consid-O erations of (1) the reactor protectica system, (2) pipe precsure drop (static and dynamic) between the point of highest pressure in the reactor coolant system and the pressurizer, (3) the pressure drop in the safety valve discharge piping, and (4) accident or transient conditions that =ay potentially cause overpressure.

Preliminary analysis indicates that the hypothetical case of vithdraval of a regulating control rod asse=bly bank frcm a relatively low power provides the basis for establishing pressurizer safety valve capacity.

The accident is teminated by high pressure reactor trip with resulting turbine trip. S is accident conditica produces a power mismatch between the reactor coolant system and stesm system larger than that caused by a turbine trip vithout i==ediate reactor trip, or by a partial load rejec-tion frcm full load.

The safety valve capacity required to prevent overprersure fcr this hypo-thetical condition is approxi=ately 600,000 lb/hr of steem. The sefaty valves are sized in accordance with the require =ents of ASME III.

435 REDUNDANCY The reactor coolant system contains two steam generators and four reactor coolant pu=ps. Operation at reduced reactor power is possible with one or more pu=ps out of service. For added reliability, power to the pumps is available frcm either of two electrically separated sources as shcvn in Figure 8-1. ,

Separate core flooding noznles are provided on opposite sides of the reac-tor vessel to insure core reflooding vater in the event of a single no:-

zie failure. Reflooding water is available frem either the core flooding -

tanks or the decay heat pu=ps which provide engineered safeguard 1cv pressure injection. The high pressure injection pipes are connected to the reactor coolant system on two of the four coolant inlet pipes.

436 SAFETY ANALYSIS The components of the reactor coolant system are interconnected by an all-velded piping system. Since the reactor inlet and outlet nozzles are lo-cated above the core, there is never any danger of the reactor coolant uncovering the core when any other system ccmponent is drained for inspec-tion or repair.

437 0FERATIONAL LIMITS Reactor coolant system heatup and cooldown rates are described in detail ,

in 4.1.4 and 4.2.6. The component stress limitations dictated by mate-rial NDIT consideraticas are described in 4.1.4 and 4 3 1.

The reactor coolant system is designed for 2,500 psig at 650 F. The nor-

=al operating conditions vill be 2,185 psig at an average system te=pera-ture of 579 F at full pcver. In this =ede of operation, the reacter ves-sel outlet temperature is 603 F. Aiditional temperature tnd pressure-

~

j variations at various power levels are shown on Figure 7-5 l

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Reactor trip signals vill be fed to the reactor protection system as a result of high reactor coolant temperature, high pressure, low pressure, and low flow, i.e., flux-flow comparator. By relating lov flow to the reactor power, opera-tion at partial power is feasible with less than four reactor ecolant pumps operating. The reactor maxi =um-calculated operating limits are as follows:

Performance Vs Pu=ps In-Service Reactor Coolant 2 Pumps 3 Pumps Operating h Pu=rs 3 Pumps (2 Loops)

Maxi =um Reactor Power,

% of Rated 100 86 60 Reactor Coolant Flov,

% of Rated 100 Th 38 Reactor operating limits under natural circulation conditions are discussed in 14.1.2.6.3. The bases for the selection of operational li=its are discussed further in 7 1.2.h.

The reactor coolant system is designed for continued operation with 1 per cent of the fuel rods in the failed condition. The tolerable radioactivity content of the coolant is based on long-term saturation activities with 1 per cent failed fuel.

h.k TESTS AND INSPECTIONS k.h.1 COMPONEaT IN-SERVICE INSPECTION Consideration has been given to the inspectability of the reactor coolant sys- 1 tem in the design of components, in the equipment layout, and in the support structures to per=it access for the purposes of inspection. Access for in-spection is defined to be access for visual examination by direct or remote means and/or by contacting vessel surfaces during nuclear unit shutdown.

k.k.l.1 Reactor Vessel Access for inspection of the reactor vessel vill be as follows:

a. Closure studs , spherical washers, and nuts can be inspected visually or by surface cont act methods.

b. External surfaces und welds on the closure head can be inspected visually or by contact following removal of the insulation. In-ternal surfaces of the closure head can be examined visually by remote means.

l O,i (f t\- 1;t, k-25 (Revised 11-6-67) Of)1]^

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l

c. Inner surfaces of the vessel outlet nozzles can be inspected 1 visually by remote = cans during refueling periods. The ecm-plete internal surface can be inspected by remote visual means following removal of the reactor core and vessel internals.

d. External surfaces of the vessel nozzle to piping velds can be inspected by remote visual means following cemoval of annulus shield plugs and vessel insulation. Insulat ion around the velds vill be made in removable sections.

e. The external surface of the reactor vessel aan be inspected during the reacter lifetime if it should br eeme necessary. A h2-in. annulus has been provided between tne reacter vessel and the primary shield to accommodate inspection equipment.

Access to this cavity can-be gained by disassembly of the shielding between the reactor vessel and pri=ary shield at the top of the annulus, which is designed to be re=ovable. The cavity has been designed to permit flooding with water to re-duce radiation levels should inspection beco=e necessary.

h.h.l.2 Pressurizer The external surface vill be accessible for surface and volumetric in-spection. The internal surface can be inspected by remote visual means.

h.h.1.3 steam Generater The external surfaces of the steam generator are accessible for surface O i and volumetric inspection. The reactor coolant side of the steam genera-tor can be inspected it ternally by remote visual means by removing the manvay covers in the steam generator heads.

Portions of the internal surface of the shell and feedvater nozzles can be inspected by re=cval of the feedvater ring, handhole covers, and man-I v87 Covers.

[ 4.k.l.h Reactor Coolant Pu=ts l

l The external surfaces of the pu=p casings are accessible for inspection.

The internal surface of the pu=p inlet is available for inspection by re-moving the pump internals.

k.4.1.5 Piping The reactor coolant piping, fittings and attachments to the piping exter-nal to the primary shield will be accessible for external surface and volumetric inspection. Reactor coclant piping velds vill not be located in the primary shielding.

h.h.l.6 Dissimilar Metal and Reeresentative Welds All dissimilar =etal velds will be made in the manufacturer's shops and vill be accessible for inspection during the service life of the nuclear unit.

L-25a (Revised 7-21-67;

, .t**

O Dissimilar metal velds on the reactor vessel include only the core flood- 1 ing lines, in-core instrumenta. tion guide tubes, and control rod drive housings.

Dissimilar metal velds in the piping include only attachments and the re-actor coolant pump inlets and outlets.

Dissimilar metal velds in the pressurizer include the surge line, the re-lief valve header, and the spray line connections.

Dissimilar metal velds on the steam generator occur caly at the small drai. lines and instrument attachments.

Representative longitudinal and circumferential velds on the piping, steam generator, pressurizer and pump casing vill be inspectable as de-scribed above. Representative velds en the reactor vessel closure head vill be inspectable. Longitudinal and circumferential veld areas on the reactor vessel interior surfaces will be inspectable.

h.h.2 REACTOR COOLANT SYSTEM TESTS AND INSPECTICNS The assembled reactor coolant system vill be tested and inspected during final nuclear unit construction and initial startup phases, as follows:

k.h.2.1 Reactor Coolant System Preeritical and Ect Leak Test This test demonstrates satisfactory preliminary operation of the entire system and its individual components, checks and evaluates operating l

(o)

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w. D000 347 h-25b (Revised 7-21-oT) l-

procedures, and deter =ines reactor coolant syste= integrit.'r at nor=al operatin6 te=perature and pressure.

4.k.2.2 Pressurizing Systen Precritical Oeerational Test This test de=enstrates satisfactory preliminary operation of the pres-surizer and its individual ec=ponents. Spray valve adjust =ents and heater centrol adjust =ents are tested, k.4.2 3 Pressurizer Surge Piping Te=perature Gradient Test The te=perature at the =idpoint of the pressurizer surge line is deter-

=ined after a period of steady state operation to check te=perature gra-dients.

4.4.2.4 Relief Syste= Test In this test all relief valves are set and adjusted, and operating proce-dures are evaluated.

4.4.2 5 Station Power S:artup Test This test deter =ines perforr.ance characteristics of the entire Station in short periods of operation at steady state pcVer levels.

4.4.2.6 Station Power Esat Balance i

This test deter =ines the actua?. reactor heat balance at various power levels to provide the necessary data for calori=etric calibration of the nuclear instrumentation and reactor coolant syste= flow rate.

4.4.2 7 Station Power Shutdown Test This test checks and evaluates the operating procedures used in shutting down the Station and detez ::ines the overall Station operating character.

istics during shutdown operaticns. These tests are in addition to the tests in co=pliance with code require =ents.

4.4 3 MAERIAL IRRADIATION SLEVEILLANCE Surveillance speci= ens of the reactor vessel shell section =aterial are installed between the core and inside vall of the vessel shell to =onitor the NDTI of the vessel =aterial during operating lifeti=e. The type of specimens included in the surveillance prc;ra= vill be Charpy V-notch l (Type A) and tensile speci= ens for =easuring the changes in =aterial prop-erties resulting frc= irradiation. This is in accordance with AS"Ji E185, l "Recc== ended Practice for Surveillance Tests en Structural Materials in Nuclear Reacters".

The reactor vessel =aterial surveillance progra= v111 utilize a total of eight specimen capsules located close to the inside reactor vessel vall g

l and directly cpposite the ce.nter portion of the cere. The locations of 1

the capsule holder tub,es'ar'e shcvn in Figures 3-45 and 3-46. In these l

'0,o 0009 348 L-26

45 REFERENCES (1) Porse, L. , Reactor Vessel Design Considering Radiatica Effects, ASME Paper No. 63-WA-LOO.

(2) Pellini, V. S. and Pu ak, P. P., Fracture Analysis Diagram Procedures for the Fracture-Safe Engineering Design of Steel Structures, Welding Research Council Bulletin 88, May 1963 (3) Robertson, T. S., Propagation of Brittle Fracture in Steel, Journal of Iron and Steel Institute, Volume 175, December 1953 (h) Kihara, H. and Masubichi, K. , Effects of Residual Stress on Brittle Fracture, Welding Journal, Volume 38, April 1959 (5) Miller, E. C., The Integrity of Reactor Pressure Vessels, ORNL-NSIC-15, May 196o.

OJ l

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positions, the spec 1= ens vill receive approximately three ti=es as =uch l radiation as the reactor vessel receives. l The =aterial from the reactor vessel vill have its initial NDTT deter-mined by the Charpy V-notch i= pact correlation with drop veight tests.

The predicted shift or change in the NDTT of the vessel =aterial result-ing from irradiation is discussed in detail in 4.1.4.1 and is shown in Figure 4-12 relative to years of operation.

~~~

The influence of neutron irradiation on the reactor vessel material prop-erties vill be evaluated periodically during Station shutdowns for re-fueling as tabulated below. l Schedule fcr Capsule Re= oval Exposure Time I Capsule Nu=ber (years from start of nuclear unit) 1 1 2 2 3 4 4 lo 5 20 6 30 7 Spares y 8 Spares This evaluation vill be acco=plished by testing samples of the material from the reactor vessel which are contained in the surveillance spec 1=en capsules. 'Ihese capsules contain steel coupons from plate, veld, and heat-affected zone material used in fabricating the reactor vessel. Do-simeters are placed with the Chaipy V-notch impr.ct specimens and tensile speci= ens. The dosimeters vill pemit evaluation of flux as seen by the specimens and vessel vall. To prevent corrosion the speci= ens are en-closed in stainless steel sheaths.

~

The irradiated samples are tested to determine the mater 1al properties, such as tensile, impact, etc., and the irradiated NDTT which may be mea-sured in a manner simile to the initial NDTT. These test results can be compared with the then-existing data on the effects of neutron flux and spectrum on engineering materials.

! The measured neutron flux mi NDTT may then be compared with the initial NUIT and the predicted NDTl shift to monitor the progress of radiation-

, induced changes in the vessel caterials. As the end of reactor design l

life nears, a significant tnerease in measured NDTT in excess of the pre-

! dicted NDTT shift could be investigated by reviewing the vessel stress .

i analysis and operating records. If necessary or required in accor.iance with the advanced knowledge available at that ti=e, the vessel transient .

limitaticas on pressure and temperature may be altered so'that vessel

( stress limits, as stated in 4.1.4 3 for heatup and cooldown, are not ex-ceeded.

l0000 349' c.- .

v 42,

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Table k-1 Tabulation of Reactor Coolar.i System Pressure Settings Ites Pressure, psig resign Pressure 2,500 Operating Pressure 2,185 Code Relief Valves 2,500 Eigh Pressure Trip 2,350 High Pressure Alarm 2,300 Low Pressure Alarn 2,150 Iow Pressure Trip 2,050 Table 4-2

's Reactor Vessel Design Cata Item Data Design / Operating Pressure, psig 2,500/2,165 Eydrotest P: essure (cold), psig 3,125 Design /OperatingTemperature,F 650/600 Overall Height of Vessel and Closure Head, ft-in. 37 4 Straight Shell Thickness, in. 8-7/16 Water Volum, ft3 4,150 Thickness of Insulation, in. 3 Number of Reactor Closure Head Stu:is 60 Flange E, in. 165 Shell D, in. 171 Inlet No::le ID, in. 28 Outlet No :le ID, in. 36 Core Flooding Water No::le E , in. 11-1/2 O .

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l Table 4-3 l Pressurizer Design Data Item Data Design / Operating Pressure, psig 2,500/2,185 Hydrotest Pressure (cold), psig 3,,125 Design /dperatingTe=perature,F 670/650 Nor=al Water Volu=e, ft) 800 Normal Steam Volume, ft3 700 Surge Line No::le Diameter, in. 10 Overall Height, ft-in. 44-0 l Table 4-4 Steam Generator Design Data O j Item Data per Ur.it DesignPressure,ReactorCoolant/ Steam,psig 2,500/1,050 i

Hyirotest Pressure (tube side-cold), Reactor Coolant, peig 3,125 Design Temperature, Reactor Coolant / Steam, F 650/600 Reactor Coolant Flow, Ib/hr 6 65 66 x 10 Heat Transferred, Btu /hr 4.21 x 109 Steam Conditions at Full Ioad, Outlet No :les:

Steam Flow, ib/hr 6 l 5 30 x 10 Steam Temperature, F 570 (35 F superheat)

Steam Pressure, psig 910

^

Feedvater Temperature, F 455 Overall Height, ft-in. 73 1/2 Shell OD, in. 147-1/k Reactor Coolant Water Volume, ft3 2,030 O

,;.. n,'" 0000 352

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O Table 4-5 Reactor Coolant Pu=p Design Data Item Data per Unit Number of Pu=ps 4 Design Pressure, psig 2,500 Hydrotest Pressure (cold), psig 3,125 Design Te=perature, F 650 OperatingSpeed(nccinal), rpm 1,180 Pumped Fluid Te=perature, F 60 to 580 Developed Head, ft 370 Capacity, gpm 88,000 O Hydraulic Efficiency, % 86 Seal Water Injection (max.), gpm 60 Seal Water Return (max.), sp:: 58 Pump Nozzle ID, in. 28 Overall Unit Height, ft 24 Water Volume, ft 95

.%) tor Stator Frsme Diameter, ft 8 Pump-Motor Moment of Inertia, lb-ft 2 70,000 Motor Data:

Type Squirrel-Cage Induction, Single Speed Voltage 6,600 Phase 3 Frequency, cps 60 Starting Across-The-I.ine Input (hot reactor ecolant), kw 5,600 Input (cold reactor coolant), kw 7,kOO

%)

nu t ,m 0000.353 k-31 1

O Table k-6 Reactor Coolant Piping Design Data Ites Data Reactor Inlet Piping ID, in. 28 Reactor Outlet Piping ID, in. 36 Pressurizer Surge Piping, in. 10 Sch. 140 l1 Design / Operating Pressure, psig 2,500/2,185 Hydrotest Pressure (cold), psig 3,125 Design / Operating Te=perature, F 650/603 Design / Operating hhperature (pressuri'zer surge line), F 670/650 Water Volu:ne, ft) 1,910 t

".able 4-7 ,

/

Transient Cycles Esti: sated Transient Description Design Cycles Actual Cycles

1. Heatup, 70 to 579 F, and Cooldevn, 579 to 70 F 480 80

'2 . Heatup, 540 to 379 F, and Cooldcun, 579 to 540 F 1,Lho 770 3 Ramp bading and Ra=p Unloading 12,000 9,000

k. Step Mading Increase 2,c00 1,500 5 Step Unicading Decrease 2,c00 1,500

6. Step Mad Reduction to Auxiliary Mad 160 12 0 7 Reactor Trip From Full Power 400 300

8. Miscellaneous Tre.nsients 10 5 3:e cycles above are based on 40-year design life. See 7 2.1.1 for ramp and step load change definitions.

• O 1 r ,s - -

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Table h-8 Design Transient Cycles Transient No. (See Table h-7) Frequency

1. Heatup, 70 to 579 F, and Cooldown, 579 to 70 F 12 per Year

2. Heatup, Sh0 to 579 F, and Cooldown, 579 to 5k0 F 36 per Year

3. Ramp Loading and Ramp Unloading 6 per Week

h. Step Loading Increase 1 per Veek 5 Step Leading Decrease 1 per Week

6. Step Loading Reduction to Auxiliary Load k per Year 7 Reactor Trip From Full Power 10 per Year Table 4-9 Reactor Coolant System Codes and Classifications Ccmponent Code Classification Reactor Vessel ASME(a) III Class A Steam Generator ASME(*) III Class A Pressurizer ASME(")'III Class A Reactor Coolant Pu=p 3 Casing ASME(") III Class A Motor IEEE,(b) NEMA,(C) and ASA(d)

Piping and Valves ASA(*) 331.1-1955 and Asso-ciated Nuclear Code Cases

• American Society of Mechanical Engineers, Bo.ler and Pressure Vessel Code.Section III covers Nuclear Vessels.

Institute of Electrical and Electronics Engineers.

(C} National Electrical Manufacturers Association.

A=erican Standards Association No. C50.2-1955 and C50.20-195h.

(e)A=er.ican Standards Association No. 331.1. ,

O I ,,,;; h-33 (Revised 11-6-67)

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O Table 4-10 Materials of Construction

__Compone:.t Section Materials Reactor Vessel Pressure Plate SA-302, Grade B Pressure Forgings A-508, (Code Case 1332)

Cladding, Stainless Weld Rod SA-371, ER 308 SA-298, E 308 Ther=al Shield and Internals SA-240, ?/pe 304 and Inconel-X Steam Generator Pressure Plate SA-232, Grade B Pressure Forgings SA-105, Grade II Cladding for Heads, Stainless SA-371, ER 308 Weld Rod SA-298, E 308 Cladding for Tubesheets Ni-Cr-Fe Tubes SB-163 (Code Case 1336)

Pressurizer Shell, Heads, and External Plate SA-212, Grade B Forgings and Nozzles SA-lOS, Grade II Cladding, Stainless Weld Rod SA-371, EH-308 SA-E98, E 308 Internal Plate SA-240, Type 304 Internal Piping SA-312, ?fpe 304 Reactor Coolant 28 in, and 36 in.

Piping Material to AS m A-212, Grade B, and A-lO6, Grade C Clad on Inner Surface to AS M A-371 Using Electrode Pfpe ER-308 10 in.

ASm A-376, ?fpe 316, and A-kO3, Grade WP 316

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O Table 4-11 References For Figure L k Increase in Transition Temperature Due to Irradiation Effects For A3C23 Steel Neutron Ref. Temp., Exposure, NDI'2, No. Reference Material Type F n[c.-j2 ( > 1 Mey) y 1 ASME Paper All Steels Max. Curve for 550. Data No. 63-a-100 (Figure 1).

2 ASTM-STP 380, A3C23 Plate Trend Curve for 550. Data p 295 3 NRL Report 6160, A3023 Plate 550 5 x 10 18 65 p 12.

4 ASTM-STP 341, Plate 550 8 x 1018 A3C23 35(a) p 226.

5 ASTM-STP 341, A3023 Plate 550 8 x 1018 100 p 226.

6 ASTM-STP 341, A3023 Plate 550 1 5 x 1019 130(*)

p 226.

7 ASTM-STP 341, A3023 Plate 550 1 5 x 1019 140 p 226.

8 Quartarly Report A3023 Plate 550 3 x 1019 12 0 of 'erogress, "Irradiatica Ef-fects on Reactor Structural Mate-rials",11-1-6h/

1-31-65 9 Quarterly Report A3023 Plate 550 3 x 1019 135 of Progress,

" Irradiation Ef-fects on Reactor Structural Mate-rials",11-1-6h/

l 1-31-65 1

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Neutron Ref. Te=p., Exposure, NIyff, No. Reference Material Type F n/c=2 (> 1 gey) p 10 quarterly Report A3023 Plate 550 3 x 1019 140 of Progress,

" Irradiation Ef-fects on Reactor Structural Mate-rials",11-164/

1-31-65 11 Quarterly Report A3T3 Plate 550 3 x 1019 170 of Progress,

" Irradiation Ef-fects on Reactor Structural Mate-rials",11-1-64/

1-31-65 12 Quarterly heport A3023 Plate 550 3 x 1019 205 of Progresa,

" Irradiation Ef- g fects on Reactor T; Structural Mate-rials, 11-1 64/

1-31-65 10 13 Welding Research A302B Weld 500 5 x 10 70 Supplement, Vol. to 27, No. 12, Oct. 575 1962, p 465-S.

14 Welding Reser.rch A302B Weld 500 5 x 1018 50 Supplement, Vol. tc, 27, No. 10, Oct. 575 1962, p 465-s.

18 15 Welding Research A3023 Weld ,500 5 x 10 37 Supplemnt, Vol. to 27, No. 1, Oct. 575 1962, p 465-S.

16 Welding Research A3C2B Weld 500 5 x 1018 25 Supplement, Vol. to 27, No. 10, Oct. 575 1962, p 465-s.

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