ML19329E137

Chapter 4 of AR Nuclear 1 PSAR, Rcs. Includes Revisions 1-18

ML19329E137
Person / Time
Site:	Arkansas Nuclear
Issue date:	11/24/1967
From:	ARKANSAS POWER & LIGHT CO.
To:
References
NUDOCS 8005300716
Download: ML19329E137 (54)

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TABLE OF CONTENTS Section Pm h REACTOR COOLANT SYSTE?4 h-1 h.1 DESIGN BASES h-1 h l.1 PERFORMANCE OBJECTIVEC h-1 4.1. 2 DESIGN CHARACTERISTICS h-1 4.1.2.1 Design Pressure h-1 h.l.2.2 Design Temperature 4-2 4.1.2.3 Reaction Loads h-2 h.l.2.h Seismic Loads h-2 L.l.2.5 Cyclic Loads h-2 h.l.2.6 Water Chemistry h-2 k.1.3 EXPECTED OPERATING CONDITIONS 4-2 L.1.4 SERVICE LIFE h-3 k.1. h .1. Material Radiation Damage h-3 k.l.h.2 Nuclear Unit Operational Thermal Cycles h-3 h l.h.3 Operating Procedures 4-h h.1.h.h Quality Manufacture 4-5 h.l.5 CODES AND CLASSIFICATIONS h-6 h.2 SYSTE4 DESCRIPTION AND OPERATION h-6 h.2.1 GE ERAL DESCRIPTION h-6 4.2.2 MAJOR COMPOIENTS h-6 h.2.2.1 Reactor Vessel h-6 h.2.2.2 Pressurizer h-7 h.2.2.3 Steam Generator h-8 h.2.2.h Reactor Coolant Pumps h-10 h.2.2.5 Reactor Coolant Piping h-11 t.

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CONTENTS (Cont'd) -m Section Pm 4.2.3 PRESSURE-RELIEVING DEVICES h-11

.4.2.k ENVIRONMENTAL PROTECTION h-11 k.2 5 . MATERIALS OF CONSTRUCTION h-12 4.2.6 MAXIMUM HEATING AND COOLING RATES h-13 h .' 2. T . LEAK DETECTION h-14 h.3 SYSTEM DESIGN EVALUATION h-15 h.3.1 SAFETY FACTORS k-15 h.3.1.1 Pressure vessel Safety 4-15 h.3.1.2 Piping 4-20 4.3.1.3 S_ team Generator h-20 h.3.2 RELIANCE ON INTERCONNECTED SYSTEMS k-21 h.3.3 SYSTEM INTEGRITY k-21 h.3.h PRESSURE RELIEF h-22 4.3.5 REDUNDANCY 4-22 4.3.6 SAFETY ANALYSIS h-22 h.3.7 OPERATIONAL LIMITS h-22 h.h TESTS AND INSPECTIONS h-23 h.h.1 COMPONENT IN-SERVICE INSPECTION h-23 k.h.1.1 Reactor Vessel 4-23 k.h.1.2 Pressurizer 4-24 h.k.1.3 Steam Generator h-2h L.h.1.k Reactor Coolant Pumps k-2h k.h.1 5 Piping k-2h h.k.1.6 Dissimilar Metal and-Representative Welds 4-2h 4.k.1.7 Inspection Schedule h-25 29i h-ii i

CONTENTS (Cent'd)

Section h

4.h.2 REACTOR COOLANT SYSTEM TESTS AND INSPECTIONS h-25 4.h.2.1 Reactor Ccolant System Precritical and Hot-Leak Test h-25 k.k.2.2 Pressurizing System Precritical Operational Test 4-25 4.h.2.3 Pressurizer Surge Piping Temoerature Gradient Test h-25 4.4.2.h Relief System Test h-25

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4.h.2.5 Unit Power Startun Test h-26 4.h.2.6 Unit Power Heat Balan_ce h-26 k.4.2.7 Unit Power Shutdown Test h-26 4.h.3 MATERIAL IRRADIATION SURVEILLANCE h-26 k.5 REFERENCES h-28 292 4-111

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LIST OF TABLES (At rear of Section)

Table No. Title h

4 Tabulation of Reactor Coolant System Pressure h-29 Settings 4-2 Reactor Vessel Design Data h-29 4-3 Pressurizer Design Data 4-30 h-4 Steam Generator Design Data h-30 L-5 Reactor Coolant Pump Design Data h-31 4-6 Reactor Coolant Piping Design Data

, h-32 4-7 Transient Cycles 4-32 k-8 Design Transient Cycles 4-33 4-9 Reactor Coolant system Codes and Classifications h-33 4-10 Materials of Construction 4-34 k-ll References for Figure 4-h -- Increase in Transition Temperature Due to Irradiation Effects for A302B Steel h i i 293 4-iv T

LIST OF FIGURES (At rear of Section)

Figure No. Title 4-1 Reactor Coolant System h-2 Reactor Coolant System Ar:tangement-Elevation 4-3 Reactor Coolant System Arrangement-Plan k-k Nil-Ductility Transition Tenperature Increase versus Integrated Neutron Exposure for A3023 Steel h-5 Reactor Vessel h-6 Pressurizer h-T - Steam Generator

-h-8 Steam Generator Heating Regions h-9 Steam Generator Heating Surface and Downcomer Level versus Power h-10 Steam Generator Temperatures h-ll Reactor Coolant Pump k-12 Predicted NDTT Shift versus Reactor Vessel Irradiation ee 294 .

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'4 REACTOR -C00li pT EYSTE*4 4.1 DESIGN BASES The reactor coolant system consists of the reactor vessel, coolant pumps, steam generators,-pressurizer, and interconnecting piping. The functional relation-ship between coolant system ecmponents is shown in Figure 4-1. The coolant sys-tem physical arrangement is shown in Figures 4-2 and 4-3.

-The reactor coolant system le designed in accordance with the following codes:

Piping.and Valves'- U3ASIB31.1-1955 (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.

Pressurizer - 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 su=marized 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 stea= generators. In addition to serving as a heat transport medium, the coolant also serves as a neutron moderator and re-flector, and as a solvent for the soluble boron utilized in chemical shim re-activity control.

As the coolant energy and radioactive material centainer, the reactor coolant system is designed to maintain its integrity under all-operating conditions.

While performing this function, the system serves the safeguard objective _of preventing the release to the reactor building of any fission products that es-cape the-primary barrier, the core cladding.

l 4.1.2 DESIGN CHARACTERISTICS

_k.l.2.1 . Design Pressure The reactor coolant system design, operating, and control set point pressures are' listed in Table h-l. The design pressure allows for operating transient

( -pressure changes. The selected design margin considers core thermal lag, cool-ant transport times-and pressure drops, instrumentation and control response 295 '

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characteristics,. and system relief valve characteristics. The design pressures s and data for the respective system components are listed in Tables 4-2 through '

4-6.

4.1.2.2 Design Temperature The design _ temperature.for each component is selected above the maximum antici-pated coolant temperature in that component.under all normal and transient load conditions. The design and operating temperatures of the respective system com-ponents are listed in Tables 4-2 through h-6.

h.1.2.3 Reaction Loads All components in the reactor coolant system are supported and interconnected l so that piping reaction forces result in ecmbined mechanical and thermal stresses in equipment nozzles and structural valls within established code limits. Equip-L ment and pipe supports are designed to absorb piping rupture reaction loads for i

elimination of secondary accident effects such as pipe motion and equipment j foundation shifting.

4.1.2.k Seismic Loads i

Reactor coolant system components are designated as Class I equipment, and are l designed to maintain their functional integrity during earthquake. The basic design guide for the' seismic analysis is the AEC publication TID-702h, " Nuclear Reactors and' Earthquake". Structures and equipment will be designed in accor-dance with Appendix 5A.

4.1.2 5 Cyclic Loads All components in the reactor. coolant system are designed to withstand the ef-fects of cyclic loads due to reactor system temperature and pressure changes.

These cyclic loads are introduced by normal unit load transients, reactor trip, and startup and shutdown operation. Design cycles are shown in Table h-7 Dur-ing unit startup and shutdown, the rates of temperature and pressure changes are limited, h.1.2.6 Water Chemistry The water chemistry is selected to provide the necessary boron content for re-activity 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 EXPECTED OPERATING CONDITIONS Throughout the load range from 15 to 100 per cent power, the reactor coolant system is operated at a constant average temperature. Reactor coolant system pressure is controlled to provide sufficient overpressure to maintain adequate core subcooling.

The minimum operating pressure is established from core thermal analysis. This analysis is based upon the maximum expected inlet and outlet temperatures, the maximum reactor power, the minimum DNER required (including instrumentation errors and the reactor control system deadband), and a core flow distribution hhm 4-2 s.

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factor. The =sximus operating pressure is established on the basis of ASME Code relief valve' characteristics and the margins required for normal pressure vari-ations in the system. Pressure control between the preset maximum and minimum

-limits is obtained directly by pressurizer spray action to suppress high pres-sure and pressurizer heater action to compensate for low pressure. Normal op-erational lifetime transient cycles are discussed in detail in h.1.4.

~4.1.4 SERVICE 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 to es-tablished operating procedures. In the following discussion each of these life-dependent. factors will be discussed with regard to the affected components.

h.1.4.1 Material Radiation Damage The reactor vessel is the only reactor coolant system component exposed to a significant level of neutron irradiation and is therefore the only component subject to material radiation damage. To assess the potential radiation dam-age at the end-of-reactor service life, the maximum exposure from fast neutrons (E > 1.0 Mev) has been computed to be 3.0 x 10 1 9 n/cm2 over a 40-year life with an 80 per cent load factor. Reactor vessel irradiation exposure calculations are described in 3.2.2.1 7 For this neutron exposure, the predicted Nil-Ductility Transition Temperature (NDTT)-shift is 250 F based on the curve shown in Figure h-4.(1) Based on an initial NDTT of 10 F, this shift would result in a predicted NDTT of 260 F.

The " Trend Curve for 550 F Data", as shown in Figure 4-h, represents irradiated material test results and was compiled from the reference documents listed in Table 4-11.

To evaluate the NDTT shift of welds, heat-affected zones, and base material 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.

h.1.h.2 Nuclear Unit Operational Thermal Cycles To establish the service life of the reactor coolant system components as re-quired by the ASME III for Class "A" vessels, the nuclear unit operating con-ditions that involve the cyclic application of loads and thermal conditions have been established for the 40-year design life.

The number of thermal and loading cycles to be used for design purposes are listed in Table h-7, " Transient Cycles". The estimated actual cycles based on a; review of ' existing nuclear- stations operations are also provided in Table

_h-7 Table 4-9 lists th6se components designed to ASME III - Class A. The effect of individual transients and the sum of these transients are evaluated to determine the fatigue usage factor _during the detail design and stress anal-'

ysis effort. As specified in ASME III Paragraph kl5.2 (d)(6), the cumulative fatigue usage factor will be less than 1.0 for the design cycles listed in Table 4-7 297 p;

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The transient' cycles listed in Table 4-7 are conservative and complete in that

-they include all-significant modes of normal and emergency operation. The es-

-~3 tbnated frequency bases for the design transient cycles are listed in Table h-8. ,

A large number of cycles of smaller magnitudes than those described can be tol-erated.

A heatup and cooldown rate of 100 F/ hour is used in the analysis of Transients 1 and 2 in Table 4-7 A ramp loading and ramp unloading transient is defined as a change in power level from 15 to 100 to 15 per cent of rated power at a rate of change of 10%/ min. A step loading transient is an instantaneous power increase or decrease of 10 per cent of rated power. A step load reduction to auxiliary load is an instantaneous reduction in electrical load from 100 to 5 per cent of rated load.

The miscellaneous transients (Item 8) listed in Table k-7 include the initial hydrotests, plus an allowance for future hydrotests in the event that reactor coolant system modifications or repairs may be required. Subsequent to a nor-mal refueling operation only the reactor vessel closure seals are hydrotested for pressure integrity; therefore, reactor coolant system hydrotests before startup are not included.

4.1.h.3 Operating Procedures The reactor coolant system-pressure vessel components are designed using the transition temperature method of minimizing the possibility of brittle fracture of the vessel materials. The various combinations of stresses are evaluated and employed to determine the system operating procedures.

The basic determination of vessel operation from cold startup and shutdown to full pressure and te=perature operation is performed in accordance with a " Frac-ture Analysis Diagram" as published by Pellini and Puzak.(2)

At temperatures below the Nil-Ductility Transition Temperature (UDTT) and the Design Transition Temperature.(DTT), which is equal to NDTT + 60 F, the pressure vessels vill be operated so that the stress levels will be restricted to a value that vill prevent brittle failure. These levels are

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

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

c. At the temperature of DTT, a maximum stress of 20 per cent yield strength.

If'the nominal stresses are held within the referenced stress limits (a through c above), brittle fracture vill not occur. This tatement is based on data re-ported by.Robertson(3) and Kihara and Masubichi(4 in published literature. It-can be shown that stress limits .can be controlled by imposing operating n oce-dures that control pressure and temperature during heatup and cooldown.(3 This procedure vill insure that the nominal stress levels do not exceed those spec-ified in a through e above.

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4.1.h.4 Quality Manufacture Material selection is discussed in detail in h.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 program vill establish the properties of the material, as received, and certify that the mechanical prop-erties of the materials in the finished vessels are consistent with those used in the design analysis. This program consists of:

a. Weld qualification test plates using production procedures and sub-jecting test plates to the heat treatments 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 sane as used for production.

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

(2) Ductility.

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

After completion of the qualification test program, production velding and in-spection procedures are prepared.

All plate or other materials are permanently identified, and the identity is maintained throughout manufacture so that each piece can be located in the fin-ished vessels. In-process and final dimensional inspections are made to insure that parts and assemblies meet the drawing requirements, and an "as-built" record of.these dimensions 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 velding elec-trodes and fluxes is tested and qualified before release to insure that required mechanical properties and as-deposited chemical properties can be met. Elec-trodes are identified and issued only on an~ approved request to insure that the correct materials are used in each veld. All velding electrodes and fluxes are maintained dry and free from contamination before use. Records are maintained and reviewed by velding engineers to insure that approved procedures and mate-rials are being used. Records are maintained for each veld joint and include the velder's name, essential veld parameters, and electrode heat or lot number.

The several types of nondestructive tests performed during vessel fabrication are as follows:

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a. Radiography, including x-ray, high voltage linear accelerator, or radioactive sources, will be used as applicable to determine the acceptability of pressure integrity velds and other velds as speci- ,)

fications require.

b. Ultrasonics is used to examine all pressure-integrity raw material and the bond between corrosion-resistant cladding to base material. In 4 addition pressure-containing velds, where applicable, are inspected by ultrasonics.

c. Magnetic Particle Examination is used to detect surface or near sur-face defects in machined veld grooves prior to velding, completed weld surfaces, closure studs, and the complete external surface of the vessels including veld seams after final heat treatment.

d. Liquid Penetrant is used to detect surface defects in the veld deposit gg i

cladding and nonmagnetic materials.

The completed reactor vessel assembly vill be shipped as a unit from the fabri-cation shop to the Plant site. The completed reactor closure head vill be shipped in like manner.

h.1 5 CODES AND CLASSIFICATIONS All pressure-containing components of the reactor coolant system are designed, fabricated, inspected, and tested to applicable codes as listed in Table h-9 h.2 SYSTEM DESCRIPTION AND OPERATION  ;

h.2.1 GENERAL DESCRIPTION The-reactor coolant system consists of the reactor vessel, two vertical once-through steam generators, four shaft-sealed coolant circulating pumps, and elec-trically heated pressurizer, and interconnecting piping. The system is arranged as two heat transport loops, each with two circulating pumps and one steam gen-erator. Reactor system design data are listed in Tables h-2 through h-6, and a system schematic diagram is shown in Figure h-l. Elevation and plan views of the arrangement of the major components are shown in Figures b-2 and h-3 4.2.2 MAJOR COMPONENTS 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 ring to which a removable reactor closure head is bolted. The reactor closure head is a spherically dished head welded to a ring flange.

The vessel has six major nozzles for reactor coolant flow, 60 control rod drive nozzles mounted on the reactor closure head, and two core flooding nozzles-- all located above the core. The reactor vessel vill be vented through the control red drives.- The vessel closure seal is formed by two concentric 0-rings with provisions between them for leak detection. The reactor vessel, nozzle design, i and seals incorporate the extensive design and fabrication experience accumu-lated by B&W Fifty-two in-core instrumentation nozzles are located en the lower head.

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The reactor closure head and the reactor vessel flange are joined by sixty 6-1/2

.in. diameter studs. Two metallic 0-rings seal the reactor vessel when the reac-tor closure head is bolted in place. Pressure taps are provided in the annulus between the two 0-rings to monitor leakage and to hydrotest the vessel closure seal after refueling.

The vessel is insulated with metallic reflective-type insulation. Insulation panels are provided for the reactor closure head.

The reactor vessel internals are designed to direct the coolant flow, support the reactor core,-and guide the control rods in the withdrawn position. The reactor vessel contains the core support assembly, upper plenum assembly, fuel asse=blies, control red assemblies , surveillance specimens , and incore instru-mentation.

The . reactor vessel shell material is protected against fast neutron flux and gam a- heating effects by a series of water annuli and the thermal shield lo-cated between the core and vessel vall. This protection is further described in 3.2.4.1.2, 4.1.4, and h.3.1.

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

Surveillance specimens made frcm reactor steel are located between the reactor vessel vall and the thermal shield. These specimens will be examined at selected intervals to evaluate reactor vessel material NDTT changes as described in h.h.3.

The reactor vessel general arrangement is shown in Figure h-5, and the general arrangement of the reactor vessel and internals is shown in Figures 3-59 and 3-60.

Reactor vessel design data are listed in Table h-2.

h.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 h-3 The electrically heated pressurizer establishes and maintains the reactor coolant pressure within prescribed limits and provides a surge chamber and a water re-serve to accom=odate reactor coolant volume changes during operation.

The pressurizer is a vertical cylindrical vessel connected to the reactor outlet piping by the surge piping. The pressurizer vessel is protected from thermal effects by a thermal sleeve on the surge line and by a distribution baffle lo-cated above the surge pipe entrance to the vessel.

Relief valves are mounted on the top of the pressurizer and function to relieve any system overpressure. Each valve has one-half the required relieving capacity.

The capacity of these valves is discussed in 4.3.h. The relief valves discharge to a quench tank located within the reactor building. The drain tank has a stored water supply to condense the steam. A relief valve protects the tank against

. overpressure.should a pressurizer valve fail to reseat.

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The pressurizer contains replaceable electric heaters in its lower section and a water spray nozzle in its upper section to maintain the steam and water at the 301 4 L

~ saturation temperature corresponding to the desired reactor coolant system pres-sure. During outsurges, as the pressure in the reactor decreases, rome of the , w; water in the pressurizer flashes to steam to maintain pressure. Electric heaters are actuated to restore the normal operating pressure. During insurges, as pressure in the reactor system increases, steam is condensed by a water spray from.the-reactor inlet lines, thus reducing pressure. Spray flow and heaters are. controlled by the pressurizer pressure controller.

Instrumentation for the pressurizer is discussed in 7.3.2.

4.2.2.3 Steam Generator The general arrangement of the steam generators is shown in Figure k-7, and de-sign data are tabulated in Table h h.

The steam generator is a vertical, straight-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 unit are the hemispherical heads, the tube-sheets, and the straight Inconel(*) tubes between the tubesheets. Tube supports hold the tubes in a uniform pattern along their length.

The shell, the outside of the tubes, and the tubesheets form the boundaries of the steam-producing section of the vessel. Within the shell, the tube bundle is surrounded by a baffle, which is divided into two sections. The upper part of the annulus between the shell and baffle is the superheater outlet, and the lower part is the feedvater inlet-heating zone. Vents, drains, instrumentation nozzles, and inspection openings are provided on the shell side of the unit.

The reactor coolant side has instrumentation connections on the top and bottom heads, manways on both heads, and a drain nozzle for the bottom head. Venting of the reactor coolant side of the unit is accomplished by a vent connection on the reactor 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 h-8 shows the flow paths and steam gen-erator heating regions.

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

a. Feedwater Heating

. Feedvater is heated to saturation temperature by direct contact heat exchange. The feedvater entering the unit is sprayed into a feed-Inconel is a trade name of an alloy manufactured by the International Nickel Company. It also has substantial common usage as a generic description of a Ni-Fe-Cr alloy conforming -to ASTM Specification SB-163. It is in the latter context that reference is made here.

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heating annulus (downcomer) formed by the shell and the baffle around I the tube bundle. The steam that heats the feedwater 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 formed by the.down-comer, the boiling sections, and the bypass steam flow to the feed-water heating region. With low (less than 1 ft/sec) saturated water velocities entering the generating section, the secondary side pres-sure drops in the boiling section are negligible. The majority of the pressure drop is due to the static head of the mixture. Conse-quently, the downcomer level of water balances the mean density of the two-phase boiling mixture in the nucleate boiling region.

b. Nucleate Boiling The saturated water enters the tube bundle, and the steam-water mix-ture flows upward on the outside of the Inconel tubes countercurrent to the reactor coolant flow. The vapor content of the mixture in- l creases almost uniformly until DNB, i.e., departure from nucleate boiling, is reached, and then film boiling and superheating occurs.

The quality at which transition from nucleate boiling to film boiling occurs is a function of pressure, heat flux, and mass velocity.

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

d. Superheated Steam Saturated steam is raised to final temperature in the superheater region.

Shown on Figure 4-9 is a plot of heating surface and downcomer level versus load.

As shown, the downcomer water level is proportional to steam 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 nucleate boiling section and the film boil-ing section is proportional to load. The surface available for.superheating varies inversely with load, i.e., as load decreases the superheat section gains from the nucleate and film boiling regions.

Mass inventory in the steam generator increases with load as the length of the heat-transfer regions varies. .

The simple concept with ideal counterflow conditions results in highly stable flow characteristics on both the reactor coolant and secondary sides. The hot reactor coolant fluid is cooled uniformly as it flows downward. The secondary side mass flow is low, and the majority of the pressure drop is due to the static j effect of the mixture. The boiling in the steam generator is somewhat similar i Lto " pool boiling", except that there is motion upward that permits some parallel l flow of water and steam. '

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A plot of. reactor coolant and steam temperatures versus reactor pcwer is shown in Figure 7-5.. As shown, both steam pressure and average reactor coolant tem-

.perature are held constant over the load range from 15 to 100 per cent rated

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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 obtain proper steam _ generator secondary mass inventory. In addition to average reactor cool-ant temperature, reactor coolant flow is also held constant. The difference between reactor coolant inlet and outlet temperatures increases proportionately as load is increased. Saturation pressure and temperature are constant, re-sulting in a variable superheater outlet temperature.

Figure 4-10, a plot of temperature versus tube length, shows the temperature differences between shell and tube throughout the steam generator at full load.

The excellent heat transfer coefficients permit the use of a secondary operating pressure and temperature sufficiently close to the reactor coolant average tem-perature so that a straight-tube design can be used.

The shell temperature is controlled by the use of direct contact steam that heats the feedvater to saturation, and the shell is bathed with saturated water from feedwater inlet to the lower 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 superheated. steam above the steam outlet nozzle, further reducing temperature differentials between tubes and shell.

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The steam generator design and stress analysis will be performed in accordance with the requirements of the ASME III as described in 4.3.1.1.

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

Shaft sealing is acco=plished 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, vould continue to oper-ate satisfactorily without seal water injection for several weeks. The out-board drain chamber would further prevent leakage to the reactor building if deterioration of the mechanical seal performance should occur.

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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 l thrust bearing is designed so that reverse rotation of the shaft vill not lead to pump or rotor damage. Lube oil cooling is accomplished by cooling coils in the motor oil reservoir. Oil pressure required for bearing lubri-cation is maintained by internal pumping provisions in the motor, or by an external system if required for " hydraulic-jacking" of the bearing surfaces for startup.

An antirotation device vill be furnished with each pump motor to prohibit reverse rotation of the pump.

Factory thrust, vibration, and seal perfor=ance 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 vill be done on subsequent units to sub-stantiate that they confom to the initial test pump characteristics.

4.2.2 5 Reactor Coolant Piping The general arrangement of the reactor coolant system piping is shown in Fig-ures 4-2 and 4-3 Piping design data are presented in Table 4-6. In addi-tion to the pressurizer surge piping connection, the piping is equipped with velded connections for pressure taps, temperature elements, vents, drains, decay heat removal, and emergency core cooling high pressure injection water.

Themal sleeves are provided in the pressurizer surge piping, and the emer-gency high pressure injection, and the core flooding connections. 3 4.2 3 PRESSURE-RELIEVDIG DEVICES The reactor coolant system is protected against overpressure by control and protective circuits such as the high pressure trip and code relief valves lo .

cated on the top head of the pressurizer. The relief valves discharge into the quench tank which condenses and collects the effluent. The schematic arrangement of the relief devices is shown in Figure 4-1. Since all sources of heat in the system, i.e., core, pressurizer heaters, and reactor coolant pumps, are interconnect'ed by the reactor coolant piping with no intervening isolation valves, all relief protection can conveniently be located on the pressurizer.

4.2.4 ENVIRONMENTAL PROTECTION The reactor coolant system is surrounded by concrete shield valls. These valls provide shielding to permit access into the reactor building for inspec-tion and maintenance of miscellaneous rotating equipment during rated power operation for periodic calibration of the incore monitoring system. These shielding valls act as missile protection for the reactor building liner plate.

Lateral bracing vill be provided near the steam generator upper tubesheet ele-vation to resist lateral loads, including those resulting from seismic forces, pipe rupture, thermal expansion, etc. Additional bracing is provided at a lower elevation to restrain the 36-in. ID vertical pipe leg from whipping.

Barriers over the reactor coolant system are also provided for shielding and missile damage protection. ,

4-11 305 5-3-68 Supplement No. 3

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4.2 5 MATERIAIS OF CONSTRUCTION Each of the materials used in the reactor coolant system has been selected for the expected environment and service conditions. The major component materials are listed in Table 4-10.

All reactor coolant system materials exposed to the coolant are corrosion-resis-tant materials consisting of 304 or 316 se, veld deposit 304 ss cladMng, Inconel (Ni-Cr-Fe), and 17-4 FH (Hn00). These materials were chosen for specific pur-poses at various locations within the system because of their superior compati-bility with the reactor coolant.

periodic analyses of the coolant chemical composition vill be perfomed to monitor the adherence of the system to the reactor coolant water quality listed in Table 9 -4 .. Maintenance of the water quality to minimize corrosion is performed by the chemical addition and sampling 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 corrosions inside the steam generators.

This required feedvater quality has been successfully used in comparable once-through, nonnuclear steem generators. 3 An external insulation of reactor coolant system components vill be compatible with the conponent materials. The reactor vessel is insulated with metallic reflective insulation on the cylindrical shell exterior. The closure flanges and the top and }

bottom heads in the area of corrosion-resistant penetrations vill be insulated with low-halide-content insulating material. All other external corrosion-resistant curfaces in the reactor coolant system vill be insulated with lov or halide-free insulating material 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 corresponding nil-ductility transition temperature (NDTT) of 10 F or less, and the material vill be tested to verify confomity to specified requirements and to detemine the actual NDIT value.

In addition, this plate vill be 100 per cent volumetrically inspected by ultrasonic test using both normal-and chcar wave.

The reactor vessel material is heat-treated specifically to obtain good notch-ductility which will insure a low NDTT and thereby give assurance that the finished vessel can be initially hydrostatically tested and operated at room temperature without, restrictions. The stress limits established for the reactor vessel are dependent upon the temperature at which the stresses are applied. As a result of fast neutron abscrption in the region of the core, the material ductility vill i change. The effect is an increase in the KDTT. The predicted end-cf-life NDTT value of the reactor vecsel opposite the core is 260 F or less. The predicted neutron exposure and EDIT shift are discussed in 4.1.4.

The unirradiated or initial NDTI of pressure vessel base plate mterial is presently measured by two methods: the drop veight test given in ASTM E208, and the Charpy V-notch inpact test (Type A) given in ASTM E23 Tne I1DIT is defined in ASIM E206

, as "the tenperature at which a specimen is broken in a series of tests 4-12 5-3-68 i

306 Supp1ement no. 3 I V

in which duplicate no-break perfomance occurs at a 10 F higher temperature".

Using the Charpy V-notch test, the IDIT is defined as the temperature at which the energy required to break the specimen is a certain " fixed" value. For SA-302B steel the ASE III Table N-332 specifies an energy value of 30 ft-lb. This value is based on a correlation with the drop veight test and vill 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 performed 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 NDTT.

The available data indicate differences as great as 40 de gees between curves plotted through the minimum and average values respectively. The determination of IMI from the average curve is considered representative of the material and is consistent with procedures specified in ASTM E23 In assessing the IMI shift due to irradiation, the translation of the average curves is used.

The material for these tests will be treated by the methods outlined in ASE 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 from the quenched edges. These tests are perfomed by the material supplier to certify the material as delivered to BW. The exact test coupon locations are reviewed and approved by BW to insure compliance with the applicable ASE Code and specifications. In accordance with ASE III Paragraphs N-313 and N-330, Bs perfoms Charpy V-notch 3 impact tests on heat-affected zone (HAZ), base metal, and weld metal on all pres-sure vessel test plates.

Differences of 20 to 40 F in HDTT have been observed between T/4 and the surface in heavy plates. The T/4 location for Charpy V-notch impact specimens is conserva-tive since material. the UDTI of the surface material is lover than that of the internal The reactor vessel design includes surveillance specimens which will pemit an evaluation of the neutron exposure-induced shift on the material nil-ductility transition temperature properties.

The remaining material in the reactor vessel and the other reactor coolant system components are purchased to the appropriate design code requirements and specific component function.

The material irradiation surveillance program is described in 4.4 3 4.2.6 bRXIMJM HEATING AND COOLING RATES The nomal reactor coolant system operating cycles are given in Tables 4-7 and 4-8 and described in 4.1.4. The nomal system heating and cooling rate is 100 F/hr.

-The exact final rates are determined during the detail design and stress m C ysis of the vessel.

The fastest cooldown rates resulting from the break of a main steam line are dis-cussed in~14.1.2 9 307 "- 3 .

5-3-68 Supplement No. 3

q 4.2 7 IEAK DETECTION

.To minimize leakage from the reactor coolant system all components are intercon-nected by an all-velded piping system. Some of the components have access open-ings of a flanged-gasketed design. The largest of these is the reactor closure head, which has a double metal 0-ring seal with provisions for monitoring for leakage between the 0-rings.

With regard to the reactor vessel, the probability of a leak occurring is con-sidered to be remote on the basis of reactor vessel design, fabrication, test, inspection, and operation at temperatures above the material IDIT as described in 4 3 1. Reactor closure head leakage vill be zero from the annulus between the metallic O-ring seals during vessel steady-state and virtually all transient operating conditions. - Only in the event of a rapid transient operation, such as an emergency cooldown, vould there be some leakage past the innemost 0-ring seal. A stress nnalysis on a similar vessel design indicates this leak rate wouldbeapproximately10cc/minthrou6h the seal monitoring taps to a drain, and no leakage vill occur past the outer 0-ring seal. The exact nature of this tran-sient condition and the resulting small leak rate vill be determined by a detailed stress analysis.

In the unlikely event that an extensive unacceptable leakage should occur from 3 the system into the reactor building during reactor operation, the leakage vill be detected by one or more of the following methods:

a. Instrumentation in t' 'l room vill indicate the addition rate of makeup wate .. .ed to maintain normal water level in the pressurizer and a the makeup cank. Deviation from normal makeup and letdown to the reactor coolant system vill provide an indication of the magnitude of the leak.

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

c. Control rocm instrumentation vill indicate the existence of a chanEe'in the veter level in the reactor building sump.

If any one of the methods above indicates unacceptable reactor coolant leakage rate 3 during operation,' the reactor vill be taken to a cold shutdown, and the cause of the problem vill be detemined.

l v 4-14 l 308 5-3-68 Supplement No. 3

h.3 SYSTEM DESIGN EVAWATI,0_N 4.3.1 SAFETY FACTORS The reactor coolant system is designed, fabricated, and erected in accordance with proven and recognized design codes and quality standards applicable for the specific component function or classification. These components are de-signed for a pressure of 2,500 psig at a nominal temperature of 650 F. The corresponding nominal operating pressure of 2,185 psig allows an adequate mar-gin for normal load changes and operating transients. The reactor system com-ponents are designed to meet the codes listed in Table L-9.

Aside from the safety factors introduced by code requirements and quality con-trol programs, as described in the following paragraphs, the reactor coolant system functional safety factors are discussed in Sections 3 and 1k.

k.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 major factors: (a) design and stress analysis, (b) quality control, (c) proper operation, and (d) relief valves.

The special care and detail used in implementing these factors in pressure vessel manufacture are briefly described as follows:

4.3.1.1.1 Design and Stress Analysis These pressure vessels are designed to the requirements of the ASME III code.

This code is a result of ten years of effort by representatives frem industry and government who are skilled in the design and fabrication of pressure ves-sels. It is a comprehensive code based on the mos' 1pplicable stress theory.

It requires a stress analysis of the entire vessel under both steady-state and transient operations. The result is a complete evaluation of both primary and 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 operations only. In establishing the fatigue life of these pressure vessels, using the design cycles from Table h-T, the fatigue evalua-tion curves of ASME III are employed.

Since ASME III requires a complete stress analysis, the designer must have at his disposal the necessary analytical tools to accomplish this. These tools are the solutions to the basic mathematical theory of elasticity equations.

In recent years the capability and use of computers have played a major part in refining these analytical solutions. The Babcock & Wilcox Company has con-firmed the theory of plates and shells by measuring strains and rotations on the large flanges of actual pressure vessels and finding them to be in agree-ment with those predicted by the theory. B&W has also conducted laboratory deflection studies of thick shell and ring combinations to define the accuracy of the theory, and is using computer programs developed on the basis of this test data.

The analytical procedure considers all process operation ecnditions. A de- i tailed design and analysis of every part of the vessel is prepared as follows:

l 309  !

h-15

1.

a. The. vessel size auul configuration are set to meet the process re- S quirements, the thickness requirements due to. pressure and other l

structural dead and live loads, and the special fillet contour and transition taper requirements at- nozzles, etc. , required by ASME III.

b. The vessel pressure and temperature design. transients given in Table 4-7_are employed in the determination 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. Comput'er programs are used in this devel-opment.-

c. The' stresses 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 distortion-free vessel).

secondary stress (to insure a vessel that vill not progressively de-form under cyclic loading), and peak stresses (to insure a vessel of maximum fatigue life).

A design report is prepared and submitted to the jurisdictional authorities and regulatory agencies, i.e., state, insurance, etc. This report defines in

- sufficient detail the design basi's,' loading conditions, etc., and will summa-

- rize.the conclusions to permit independent checking by interested parties.

• k.3.1.1.2 Quality control In-process and final dimensional inspections are made to insure that parts and assemblies . meet the drawing requirements, and an "as-built" record of these dimensions is kept for reference. A temperature-controlled gage-room is main-L tained to keep all measuring equipment in proper calibration, and personnel supervising this work are trained in formal programs sponsored by gage equip-ment manufacturers.

The' practice of applied radiography is being continually improved to enhance flaw detection; Present procedures are:

a. All velds are properly prepared by chipping and grinding valleys be-tween stringer beads .so that radiographs can be properly interpreted,

b. lAll radiographs are reviewed by two people knowledgeable and skilled '

.in their interpretation.

.c.

An 0.080-in. lead filter is used at the film to absorb " broad-beam seatter" when usin'g'high-voltage equipment (above 1 Mev).

d. ' Fine grain or extra fine grain film is used for all exposures.

e. Densities of radiographs are controlled by densitometers. ,

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f.  : Double film technique is used on all gamma-ray exposures as well as high voltage exposures.

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g. Films are processed through an automatic processor which has a con-trolled replenishment, temperature, and process cycle, all contribut-ing to better quality.

h. Energies .are controlled so as to be in the optimum range.

Ultrasenics, one of the most useful of inspection tools, is being used as fol-lows:

s. In addition to radiography, pressure-containing velds, where applicable, h are inspected by ultrasonics.

b. In order to detect laminations which are normally parallel to the surface, plates are also inspected by a shear wave.

c. The bond ber.veen cladding and base material is inspected by ultra-sonics.

d. All plate is 100 per cent volumetrically inspected by ultrasonics 3 using both normal and shear wave,

e. Personnel conducting ultrasonic inspections are given extensive train-ing.

The magnetic particle examination is used to aid in detecting surface and near surface defects and Lis employed on both parts and finished vessels as follows:

a. Welds are inspected with the magnetic particle method after removal

, of backup strips.

b. Weld preparations are inspected by the magnetic particle method.

c. The external surface of the entire vessel, including veld seams, is inspected after all heat treatment.

d. Reactor vessel closure studs. 5

e. Personnel using this method are trained by B&W and by the equipment manufacturers who offer formal training programs.

The liquid penetrant examination helps to detect surface defects and is partic-ularly adaptable to the nenmagnetic materials such as stainless steel. It is presently being used as follows:

a. Inspection of weld-deposited cladding.

b. Personnel using this method of examination are trained by B&W and by 5 the' equipment manufacturers.

. The primary purpose of these quality control procedures and methods is to lo- 1 cate, . define,- and determine the size of material defects .to allow an evaluation j of defect acceptance, rejection, or repair. '

Ac-i a

}{j T-h-68 Supplement No. 5 h-lT _._ _ __ . - -l

The size cf defcet that can conesiv:bly contribute to the rupture of a versel depends not cnly on the siza effect but en the cri:ntat' ion Cf the defact, the magnitude of the stress field, and temperature. These major parameters have been c:rrelated by Pellini and Puzak(2) who have prepared a " Fracture Analysis Diagram" q vhich is the basis of vessel operation from cold startup and shutJwn to full

• Pr;ssure and temperature operation.

The diagram predicts that, for a given level of stress, larger flav sizes vill be rsquired for fracture initiation above the NDTT temperature. For example, at stresses in the order of 3/4 yield strength, a flaw in the order of 8 to 10 in.

may be sufficient to initiate fracture at temperatures below the NDTT temperature.

However, at NDTT +30 F, a flav of 1-1/2 times this size may be required for ini-tiation of fracture. While at a temperature of NDTT +60 F, brittle fracture is not possible under elastic stresses because brittle fracture propagation does not take place at this temperature. Fractures above this temperature are of the pre-dominantly ductile type and are dependent upon the member net section area and saction modulus as they establish the applied stress.

Stud forgings vill be inspected for flaws by two ultrasonic inspections. An axial longitudinal beam inspe:: tion vill be perfomed. The rejection standard vill be less-of-back-reflectiongreaterthanthatfroma1/2-in.diameterflatbottom hole. A radial inspection vill be made using the longitudinal beam technique.

This inspection vill carry the same rejection standards as the axial inspection.

In addition to the ultrasonic tests, liquid-penetrant inspection vill be perfomed cn 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 th) 40-year design life. Therefore, there vill be no planned frequency for stud raplacement. 3 3

Ten per cent of the studs vill be given a thorough inspection at every refueling.

This inspection vill include visual, dye-penetrant, and ultrasonic examination 3 procedures. Any stud showing indication of a defect vill be replaced.

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

diameter studs. The stud material is A-540, Grade B23 (ASME III, Case 1335) which has a minimum yield strength of 130,000 psi. The studs, when tightened for oper-ating conditions, vill have a tensile stress of approximately 30,000 psi. Thus, at operatin6 conditions (2,185 psig):

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

b.

25 adjacent studs can fail before the remaining studs reach yield strength. -

c. 26 adjacent scuds can fail before the remaining studs reach the ultimate tensile strength,

d. 43 symmetrically Jocated studs can fail before the remaining studs reach yield strength.

l l

4-18 l 1 l '312 5-3-68 1 Supplement No. 3

h.3.1.1 3 operation As previously mentioned in h.1.h, pressure vessel service life is dependent on adherence to established operating procedures. Prescure vessel safety is also dependent on proper vessel operation. Therefore, particular attention is given to fatigue evaluation of the pressure vessels and to the factors that affect fatigue life. The fatigue criteria of ASME III are the bases of designing for fatigue. They are based en fatigue tests of pressure vessels sponsored by the AEC and the Pressure Vessel Research Cc=mittee. The stress limits established for the pressure vessels 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 reactor vessel material ductility vill change. The effect is an increase in the nil-ductility transition te=perature (ULTT). The determination of the predicted NDTT shift is described in h.1.h.1. This NDTT shift is factored into the plant startup and shutdevn procedures so that full cperating pressure is not attained until the reactor vessel temperature is abcVe the design transition temperature (DTT). Belev the DTT the total stress in the vessel vall due to both pressure and the associated heatup and cocidown transient is restricted to 5,000 - 10,000 psi, which is below the threshcld of concern for safe operation. These stress levels define an operating coolant pressure temperature path or envelope for a stated heatup or cooldown rate that must be folleved. Additional infernation on the determination of the operating procedures is provided in h.1.k.1, h.1.h.2, and h.1.h.3.

i h.3.1.1.k Additional Pressure Vessel Safety Factors Additional methods and precedures used in pressure vessel design, not previ-ously mentioned in h.3.1.1 above but which are considered conservative and pro-vide an additional margin of safety, are us follevs:

a. Use of a stress concentration factor of 4 on assumed flaws in calcu-lating stresses,

b. Use of minimum specified yield strength of the material instead of the actual values.

c. Neglecting the increase in yield strength resulting frem irradiation effects. '

d. The design shift in HDTT as given in h.1.h.1 is based on maximum pre-dicted flux levels at the reactor vessel inside vall surface, where-as the bulk of the reactor vessel material vill experience a lesser exposure of radiation and consequently a lower 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 vill be sub-jected to a greater degree of irradiation and therefore to a larger.

shift in NDTT value than vill be experienced by the vessel. The 3 specimens lead the vessel with respect to irradiation effects and im- l

- part a degree of conservatism in the evaluation of the capsule spec-imens. The material irradiation surveillance program is described in h.h.3.

t" h-19 513

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 assumptions, uncertainties, and comparisons of calculational 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, fabri-cation, inspection, and operating procedures used to insure confidence in the in-tegrity of pressure vessels. Experience reported by Reference 5, a s , and the satisfactory experience of BE custcuners support the conclusion that pressure ves-sel rupture is-incredible.

4 3 1.2 Piping Total stresses resulting from themal expansion and pressure and mechanical and seismic loadings are all considered in the desi6 n of the reactor coolant piping.

The total stresses that can be expected in the piping are within the maximum code allowables. The pressurizer surge line connection and the high pressure injection connections are equipped with thermal sleeves to limit stresses from themal shock to acceptable values. All materials and fabrication procedures vill meet the re-quirements of the specified code. All material vill be ultrasonically inspected.

All velds vill be radiographically inspected. All interior surfaces of the inter-connecting piping are clad with stainless steel to eliminate corrosion problems and to reduce coolant contamination.

4313 steam Generator Because the basic concept of the once-through steam generator vould indicate the possible existence of differential thermal expansion-induced stresses in either the tubes or shell, the themal loadings have been evaluated using the most severe design transients from Table 4-7 The basic structural premise of the steam generator is that the tubesheets them-selves are designed to take the full design pressure on either side of the tube-sheet with zero pressure on the other side. That is, the tubes are not counted upon for any structural aid or support.

The steam line failure analyzed in 1L.1.2 9 closely simulates this design premise in a transient manner. Secondary temperature variations during the accident are essentially transient skin effects with the controlling temperature for the tube-cheets and tubes being that of the reactor coolant. Themal stresses for this case vill be below ASME allowable values. Some tube deformation may occur but will be restrained by the ttice supports.

During nomal power operation the tubes are hotter than the shell of the steam generator by 10 to 20 F depending on load. The effect is to put the tubes in a slight cocpression of 3,000 psi 80 the 20 F maximum temperature difference.

During startup and shutdown operations the tubes are hotter than the shell of the steam generator by 40 F. This places the tubes in a cocipressive stress of 6,000 psi. Thus, the stress levels developed during normal startup or shutdown opera-tions :ause no adverse effect on the tubes since these stresses are well below the allovable stress of 23,3c0 psi fer SB-163 material. Buckling of the tubes 3 does notLoccur sinee they are supported 314 4 .

5-3-68

, Supplement No. 3

lateralhr at LO-in. intervals along their length. To demonstrate the structural adequacy of the steam generator at this condition, a laboratory uait was con-structed of the same tube size, length, and material 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 perfomed with a tube-to-shell temperature differance twice as great as the maximum expected during startup and shutdown (Transients 1 and 2, Table 4-7). Destructive e-htion of the unit after this test indicated no adverse effects from fatigue, stress, buckling, or tube-to-tube-sheet joint leakage.

4.3.2 RELIANCE ON INTERCONNECTED SYSTEMS The principal heat re:mval systems vnu: 1 ara inter:onnected with the reactor coolant system.

system are the steam and feedwater systems and the decay heat removal The reactor coolant system is depepdent upon the steam generators, and the steam, feedvater, and condensate systems for decay heat removal from normal operating conditions to a reactor coolant te=perature of approximately 250 F.

All vital active co=ponents in these systems are duplicated for reliability purposes.

The enginet. ring flow diagram of the steam and feedwater 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 atored in the feedwater system may be phere. into the steam generators and the resultant steam vented to the atmos-pumped A motor driven, 5 per cent capacity emergency feed pa=p will supply water to the steam generators when power is available. A 5 per cent capacity turbine driven emergency feed pump will serve as back up in the event of loss of all electrical power. The decay heat removal system is used to remove decay heat when the themal driving head of the reactor coolant system is no longer adequate to generate steam. This system is completely described in 9.5. The heat received by this system is ultimately rejected to the service system which also contains sufficient redundancy to guarantee proper operation.

A schematic diagram of the service water system is presented in Figure 9-4 4.3.3 SYSTEM INTEGRITY The integrity of the reactor coolant system is insured by proper materials selection, fabrication quality control, design, and operation. All com-ponents in the reactor coolant system are fabricated from materials initially having a low nil-ductility transition te=perature (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 reactor vessel, provisions are included for materials surveillance specimens. These will be periodically examined, and any required temperature-pressure restrictions will be incorporated into reactor operation to insure operation above NDTT.

The reactor coolant system is designed in accordance with ASME pressure vessel and ASA power piping codes as covered in 4.1. Relief valvea on the preasurizer are thansized to cent.

10 per prevent system pressure from exceeding the design point by more As a further assurance of system integrity, all components in the system will i

' be hydrotested at 3,125 psig before initial operation. The largest and most frequently used opening in the reactor coolant system,'i.e., the reactor closure 4-21 2-8-68 Amendment No. 1

hiad,-contains provisions for separate hydrostatic pressurization between the 0-rinE type gaskets.

k.'3.h PRESSURE RELIEF

-The reactor coolant system is protected against overpressure by safety valves located on the top of the pressurizer.

The' capacity of the pressurizer safety valves is determined from considerations

~

of (a) the reactor protection system, (b) pipe pressure drop (static and dynamic) between the point of highest pressure in the reactor coolant system.and the pressurizer, (c) the pressure drop in the Pafety. valve discharge pipfng, and (d) accident or transient conditions that may pv entially cause overpressure.

Preliminary analysis indicates that the hypothetical case of withdrawal of a regulating control rod assembly bank from a relatively low power provides the basis for establishing pressurizer safety valve capacity. The accident is ter-minated.by high pressure reactor trip with resulting turbine trip. This acci-dent condition produces a power mismatch between the reactor coolant system and steam system larger than that caused by a turbine trip without immediate reactor trip, or by a partial load rejection from full load.

The ASME Section III required safety valve capacity as determined on the basis 3/ 5 of the accident described above is 600,000'1b/hr. Two 300,000 lb/br valves are installed. An additional pilot-operated safety valve, capable of 100,000 lb/hr, is provided to limit the lifting frequency of the code safety valves.

h.3 5 REDUNDANCY

)

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

Separate core flooding nozzles are provided on opposite sides of the reactor ves-sel to insure core reflooding water in the event of a single nozzle failure. Re-flooding water is available from either the core flooding tanks or the decay heat pumps which provide engineered safe 6uard low pressure injectir n. The high pressure 3 injection pipes are connected to the reactor coolant system ot each of the four coolant inlet. pipes.

k.3.6 -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 located above the core, there is never any danger of the reactor coolant uncovering the core when cny other system component is drained for inspection or repair.

i I

h.3.T '0FERATIONAL LIMITS

! Reactor coolant system heatup and cooldown ratas are described in detail in h.l.h

.and k.2.6. The component stress limitations dictated by material NDTT censidera-

-tions are described in h.1.h and 4.3.1.

3-68 3 } () Supplement No. 5 h

The reactor coolant system is designed for 2,500 psig at 650 F. The nomal operating conditions v111 be 2,185 psig at an average system temperature of 579 F at rated power. In this mode of operation, the reactor vessel outlet temper-ature is 603 F. Addi,ional temperature variations at various power levels are shown on Figure 7-5 Reactor trip signals vill be fed to the reactor protection system as a result of high reactor coolant temperature, high pressure, low pressure, lov flow, and 3 number of operating pumps. By relating lov flow to the reactor power, operation at partial power is feasible with less than four reactor coolant pumps operating.

The reactor maximum-calculated operating limits are as follows:

Perfomance Vs Pumps In-Service Reactor Coolant 2 Pumps Pumps Operating 4 Pumps 3 Pumps (2 Loops)

Maximum Reactor Power,

% of Rated 100 86 60 Reactor Coolant Flow,

%ofRated 100 74 38 Reactor operating limits under natural circulation conditions are discussed in 14.1.2.6 3 The bases for the selection of operational limits are discussed further in 7 1.2.4.

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-tem saturation activities with 1 per cent failed fuel (Table 11-3).

4.4 TESTS AND IIGPECTICIE 4.4.1 COMPONENT DT-SERVICE IIGPECTION Consideration has been given to the inspectability of the reactor coolant system in the design of components, in the equipment layout, and in the support struc-tures to pemit access for the purposes of inspection. Access for inspection is defined to be access for visual examination by direct or remote means and/or by contacting vessel surfaces during nuclear unit shutdown.

4.4.1.1 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 contact methods. I

b. External surfaces and welds on the closure hesd can be in-spected visually or by contact following removal of the in-sulation. Internal surfaces of the closure head can be exadned visually by remote means.

s 4-23 .

M7 5-3-68 Supplement No. 3

- = -

0.

Inner surfaces of the vessel outlet no::les can be inspected visually "'3 by remote means during refueling periods. The complete internal sur-face can be inspected by remote visual means following removal of the reactor core an.1 vessel internals.

d. External surfaces of tne vessel no::le to piping velds can be in-

.spected by remote visual means following removal of annulus shield plugs and vessel insulation.

Insulation around the velds vill be made in removable sections,

e. The external surface of the reactor vessel can be inspected during the reactor lifetime if it should become necessary. An annulus has been provided between the reacter vessel and the primary shield to accommodate inspection equipment. Access to this cavity which can 6 be flooded can be gained by disassembly of the shielding between the reactor vessel and primary shield at the top of the annulus, which is designed to be removable.

L.L.l.2 Pressurizer The external surface vill be accessible for surface and volumetric inspection.

The internal surface can be inspected by remote visual means.

k.L.l.3 Steam Generater The external surfaces of the steam generator are accessible for surface and 3

volumetric inspection. The reactor coolant side of the steam generator can be inspected internally by remote visual means by removing the manway covers in the steam generator heads.

Portions of the internal surface of the shell and feedvater no::les can be in-spected by removal of ne feedvater ring, handhole covers , and manvay covers.

i L.L.l.L Heactor Coolant Pu=cs The external surfaces of the pump easings are accessible for inspection. The internal surface cf the purp :nlet ir available for inspection by removing the pump internals.

L.4.1.5 Piping The reactor coplant pipias, fittir.gs and attawhments to the piping external to the primary shield vill be accessible for external surface and volumetric in-spection. Reactor coolant piping veles vill nct be located in the primary shielding.

h.h.l.6 Dissimilar Metal and Representative 'n' elds All dissimilar metal velds vill ce made in the manufacturer's snops and vill be accessible for inspection Guring the service life of the nuclear unit, b\

h-24 7-11-68 Supplement No. 6

Dissimilar metal velds on the reactor vessel include only the core flooding lines, in-core instrumentation guide tubes, and control rod drive housings.

Dissimilar metal velds in the piping include only attachments and the reactor coolant pump inlets and outlets.

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

Dissimilar metal velds on the steam generator occur only at the small drain lines and instrument attachments.

Representative longitudinal and circumferential velds on the piping, steam gen-erator, pressurizer, and pump casing vill be inspectable as described above. ,

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

Longitudinal and circumferential veld areas on the reactor vessel interior sur-faces vill be inspectable, b.k.1 7 Inspection Schedule The schedule for the type and frequency of inspection in each of the areas mem-tiened above vill be established during the detailed design.

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

k.4.2.1 Reactor Coolant System Precritical and Hot Leak Test This test demonstrates satisfactory preliminary operation of the entire system and its individual components, checks and evaluates operating procedures, and determines reactor coolant system integrity at normal operating temperature and pressure.

4.k.2.2 Pressurizing System Precritical Operational Test This test demonstrates satisfactory preliminary opere. tion of the pressurizer and its individual components. Spray valve adjustments and heater control ad-justments are tested.

h.4.2.3 Pressurizer Surge Piping Temperature Gradient Test The temperature at the midpoint of the pressurizer surge line is determined after a period of steady-state operatton to check temperature gradients.

4.h.2.h Relief System Test In this test all relief valves are set and adjusted, and operating procedures are evaluated.

4 3}9 h-25

h.h.2 5 Unit Power Startup Test

,)

This test determines performance characteristics of the entire unit in short '

periods of operation at steady-state power levels.

h.h.2.6 Unit Power Heat Balance This test determines the actual reactor heat balance at various power levels to provide the necessary data for calorimetric calibration of the nuclear instrumentation and reactor coolant system flow rate.

h.h.2.7 Unit Power Shutdown Test This test checks and evaluates the operating procedures used in shutting down the unit and determines the overall unit operating characteristics during shutdown operations. These tests are in addition to the tests in compliance with code requirements.

h.h.3 MATERIAL IRRADIATION SURVEILLANCE Surveillance specimens of the reactor vessel shell section material are in-stalled between the core and inside vall of the vessel shell to monitor the NDTT of the vessel material during operating lifetime. The type of specimens b included in the surveillance program vill be Charpy V-notch (Type A) and ten-sile specimens for measuring the changes in material properties resulting from irradiation. This is in accordance with ASTM E185-66, " Recommended Practice for Surveillance Tests on Structural Materials in Nuclear Reactors."

The reactor vessel material surveillance program vill utilize a total of four 5 surveillance specimen holder tubes located close to the inside reactor vessel vall, The locations of the capsule holder tubes are shown in Figures 3-59 and 3-60. In these positions,the specimens vill receive approximately three times as much radiation as the reactor vessel receives. ~

l5 As recommended in ASTM E185-66, paragraph h.6, and indicated in the table below, i specimens vill be withdrawn at three or more separate times, and one of the 15 data points obtained shall correspond to the neutron exposure of the vessel vall near the end of its design life. As can be seen from Figure h-12, the planned specimen removal prcgram vill provide sufficient data points (indicated on curve) to allow construction of the curve of NDTT shift versus integrated neutron exposure for the Russellville reactor pressure vessel.

The material from the reactor vessel vill have its initial NDTT determined by the Charpy V-notch impact correlation with drop veight tests. The predicted shift or change in the NDTT of the vessel material resulting from irradiation is discussed in detail in h.l.h.1.

The influence of neutron irradiation on the reactor vessel material properties will be evaluated periodically during unit shutdowns for refueling. Adequate specimen holders are provided to permit evaluation on approximately the schedule shown below.

320 7-3-68 k-26 Supplement No. 5

.3 Schedule for Causule Removal Equivalent Vessel Material Exposure Time Holder Tubes (Years) 5 1 10 2 20 3 30

~h h5 4 This evaluation vill be accomplished by testing samples of the material from the reactor vessel which are contained in the surveillance specimen capsules.

These capsules contain steel' coupons from plate, veld, and heat-affected zone material used in fabricating the reactor vessel. Dosimeters are placed with the Charpy V-notch' impact specimens and tensile specimens. The dosimeters vill permit evaluation of-flux as seen by the specimens and vessel vall. To prevent corrosion the specimens are enclosed in stainless steel sheaths.

The irradiated samples are tested to determine the material properties, such as tensile, impact, etc., and the irradiated NDTT which may be meausred in a manner similar to the initial NDTT. These test results can be compared with the then-existing data on the effects of neutron flux and spectrum on engineer-ing materiale.

The measured neutron flux and NDTT may then be compared with the initial UDTT and the preducted NDTT shift to monitor the progress of radiation-induced changes in the vessel materials. As the end of reactor decign life nears, a significant increase in measured NDTT in excess of the predicted NDTT shift could be investigated by reviewing the vessel stress analysis and operating records. If necessary or required in accordance with the advanced km:vledge available at that time, the vessel transient limitations on pressure and tem-perature may be altere? To that vessel stress limits, as stated in h.1.4.3 for heatup and cooldown, art not exceeded, k

32 l- -T-3-68 Supplement No. 5 W

1 I

k.5 REFERENCES (1) Porse, L., Reactor Vessel Design Considering Radiation Effects, ASME Paper No. 63-WA-100.

(2) Pellini, W. S. and Puzak, P. P., Fracture Analysis Diagram Procedures for the Fracture-Safe Engineering Design of Steel Structures, Welding Research Council Bulletin 88, May 1963. l (3) Robertson, T. S. , Propagation of Brittle Fracture in Steel, Journal of Iron and Steel Institute, Volume 175, December 1953.

(4) Kihara, H. and Masubichi, K., Effects of Residual Stress on Brittle Frac-ture, Welding Journal, Volume 38, April 1959 (5) Miller, E. C., The Integrity of Reactor Pressure Vessels, ORNL-NSIC-15, May 1966.

322 u-28

Table 4-1 Tabulatien of Reactor Coolant System Pressure Settings Item Pressure rsig Design Pressure 2,500 Operating Pressure 2,185 Code Relief Valves 2,500 High Pressure Trip 2,350 High Pressure Alarm 2,300 Lov Pressure Alarm 2,150 Low Pressure Trip 2,050 Table 4-2 Reactor Vessel Design Data Item Data Design / Operating Pressure, psig 2,500/2,185 Hydrotest Pressure (cold), psis 3,125 Design / Operating Temperature, F 650/600 Overall Height of Vessel and Closure Head, ft-in. 37 h Straight Shell Thickness, in. 8-7/16 Water Volume, ft3 h,150 Thickness _ of Insulation, in. 3 Number of Reactor Closure Head Studs 60 Flange, ID, in. 165 Shell ID, in. 171 Inlet Nozzle ID, in. 28 Outlet Nozzle ID,.in. 36 Core Flooding Water Nozzle ID, in. 11-1/2

} TABLES 4-1, 4-2 4-29 1

Table 4-3 Pressurizer Design Data O Item Dat a

~ Design / Operating Pressure, psig 2,500/2,185 Hydrotest Pressure (cold), psig 3,125 Design / Operating Temperature, F 670/650 Normal Water Volume, ft3 goo Normal Steam Volume, ft3 700 Surge Line Nozzle Diameter, in. 10 Overall Height, ft-in. h4-0 Table h h Steam Generator Design Data Item Data per Unit Design Pressure, Reactor Coolant / Steam, psig 2,500/1,050 Hydrotest Pressure (tube side-cold, reactor coolant), psig 3,125 Design Temperature, Reactor Coolant / Steam, F 650/600

- Reactor Coolant Flow, lb/hr 65.66 x 106 Heat Transferred, Btu /hr h.21 x 109 Steam Conditions at Rated Load, Outlet Nozzles:

Steam Flov, lb/hr' 5.30 x 10 6 Steam Temperature, F 570 (35 F superheat)

Steam Pressure, psig 910 Feedwater Temperature, F 455 Overall Height, ft-in. 73 1/2 Shell OD, in. 147-1/h Reactor Coolant Water Volume, ft3 2,030 TABLES h-3, h b h-30 ,

^

324 A. :

Table 4-5 Reactor Coolant Pump Design Data Item Data per Unit Number of Pumps k Design Pressure, psig 2,500 Hydrotest Pressure (cold), psig 3,125 Design Temperature, F 650 Operating Speed (nominal), rpm 1,168 Pu= ped Fluid Temperature, F 60 to 580 Developed Head, ft 362 Capacity, gpm 88,000 Hydraulic Efficiency, 7 86 Seal Water Injection (max), gpm 50 Seal Water Return (max), gpm 58 Pu=p Nozzle ID, in. 28 Overall Unit Height, ft 2h Water Volume, ft3 95 Motor Stator Frame Diameter, ft 7 - 1/2 Pump-Motor Moment of Inertia, Ib-ft2 70,000 Motor Data:

Type Squirrel-Cage Induction, Single Speed Voltage 6,600 Phase 3 Frequency, cps 60  ;

Starting Across-The-Line Input (hot reactor coolant), kw 5,600 Input (cold reactor coolant), kw T,400

}

TABLE h-5 h-31 REVISED, 2-8-68 L_ }

Table k-6 Reactor Coolant Piping Design Data

)

Item Data Reactor Inlet Piping ID, in. 28 Reactor Outlet Piping ID, in. 36 Pressurizer Surge Piping, in. 10 Sch. 1k0 Design / Operating Pressure, psig 2,500/2,185 Hydrotest Pressure (cold), psig 3,125 Design / Operating Temperature, F 650/603 Design / Operating Temperature (pressurizer surge line), F 670/650 t

Water Volume, ft3 1,91o t

Table h-7 Transient Cycles ,

(Rated Steam Load Basis)

Estimated Transient Description Design Cycles Actual Cycles

1. Heatup, 70 to 579 F, and Cooldown, 579 to 70 F h80 80

2. Heatup, 5ho to 579 F, and Cooldown, 579 to 540 F 1,hh0 770

3. Ramp Loading and Ramp Unloading (15-100-15%) 12,000 9,000

h. Step Loading Increase (10%) 2,000 1,500 5 Step Unloading Decrease (10%) 2,000 1,500

6. Step Load Reduction to Auxiliary Load (100-50%) 160 120 7 Reactor Trip from Rated Power 400 300

8. Miscellaneous Transients 10 5 The cycles above are based on 40-year design life.

- TABLES h-6, h-7 326 h-32 i E

Table b-8 Design Transient Cycles (Rated Steam Load Basis)

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 540 F 36 per Year

3. Ramp Loading and Ranp Unloading (15-100-15%) 6 per Week

h. Step Leading Increase (10%) 1 per Week 5 Step Unloading Decrease (10".) 1 per Week

6. Step Loading Reduction to Auxiliary Load (100-5%) h per Year 7 Reactor Trip from Rated Power 10 per Year Table h-9 Reactor Coolant System Codes and Classifications Ccmponent Code Classification Reactor Vessel ASME(a) III Class A Steam Generator ASME(a) III Class A Pressurizer ASME(") III Class A Reactor Coolant Pump Casing ASME(*) III Class A Motor IEEE,(b) NEMA,(C) and USASI(d)

Piping and Valves USASII *) B31.1-1955 and Asso-ciated Nuclear Coda Cases (a)American Society of Mechanical Engineers, Boiler and Pressure Vessel Code.Section III covers Nuclear Vessels.

(b) Institute of Electrical and Electronics Engineers.

(c) National Electrical Manufacturers Association.

d) United States of America Standards. Institute Nos. C50.2-1955 (e) United States of America Standards Institute No. B31.1.

}y TABLES 4-8, h-9 h-33

Table h . Materials of Construction '

Component Section Material Reactor Vessel Pressure Plate SA-533, Grade B Class 1 (*)

Pressure Forgings A-508-6h Classes 1 & 2 Code Case 1332-3 Cladding 18-8 Stainless Steel Thermal Shield and Internals SA-2h0, Type 304, and Inconel-X Steam Generator Pressure Plate SA516 Grade TO SA533 Grade B Class 1 Pressure Forgings A-508-64 Class 1 (Code Case 1332-3)

Cladding for Heads 18-8 Stainless Steel Cladding for Tubesheets Ni-Cr-Fe ,

Tubes SB 163 '

Pressurizer Shell, Heads, and External Plate SA516 Grade 70 i Forgings A508 (Code Case 1332-3) -

Class 1 Cladding 18-8 Stainless Steel Internal Plate SA-2ho, Type 30h Internal Piping SA-312, Type 30h Reactor Coolant 28 in. and 36 in. SA516 Grade 70 (Elbovs)

Piping A106, Grade C (Straight)

Cladding 18-8 Stainless Steel 10 in.

A376, Type 316, (Straight)

Ah03, Grade WP-316 (Elbows)

(*)' This material'.is metallurgically identical'to SA-302, Grade B,,as modified by Code Case 1339 ,

4-34 ,

e+ . 5-3-68

) ,

Supplement No. 3 m . m

1 Table b-11 References for Figure h-h -- Increase in Transition Temperature Due to Irradiation Effects for A302B Steel Neutron Ref. Temp., Exposure, NDTT, No. Reference Material Type F n/cm2 (>l Mev) F 1 ASME Paper All Steels Max. Curve for 550 Data No. 63-WA-100 (Figure 1).

2 ASTM-STP 380, A302B Plate Trend Curve for 550 Data P 295 3 NRL Report 6160, A302B Plate 550 5 x 10 65 p 12 h ASTM-STP 3hl, A302B Plate 550 8 x 10 10 85f "

p 226 5 ASTM-STP 3hl, A302B Plate 550 8 x 10 10 100 p 226 6 ASTM-STP 341, A302B Plate 550 1 5 x 10 19 130(*)

p 226 7 ASTM-STP 3hl, A302B Plate ~ 550 1.5 x 10 19 1ho p 226 8 Quarterly Report A302B Plate 550 3.x 10 19 120 of. Progress,

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

1-31-65 9 Quarterly Report A302B Plate 550 3 x 10 19 135 on Progress,

" Irradiation Ef-fects on Reactor

' Structural Mate-rials," 11-1-6h/

1-31-65

" Transverse specimens.

J TABLE h-11 329 h-35 t

1 Table h-11 (Cont'd)

Neutron Ref.- Temp., Exposure, NDTT,

-No. Reference Material Type F n/cm2 (>l Mev) F 10 Quarterly Report A302B Plate 550 3 x 10 19 140 of Progress,

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

1-31-65 11 Quarterly Report A302B Plate 550 3 x 10 19 170 of Progress,

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

1-31-65

'12 Quarterly Report A302B. Plate 550 3 x 10 19 205 of Progress,

" Irradiation Ef-fects on Reactor Structural Mate- -

rials," 11-1-6h/

1-31-65 13 Welding Research A302B Weld 10 500 5 x 10 70 Supplement, Vol. to' 27, No. 10, Oct. 575 1962, p 465-S 1h Welding Research Weld 10 A302B 500 5 x 10 50 Supplement, Vol. to 27, No. 10, Oct. 575 1962, p h65-S 15 Welding Research A302B Weld 10 500 5 x 10 37 Supplement, Vol. to 27, No. 10, Oct. 575 1962, p 465-S 16 Welding Research A302B Weld 10 500 5 x 10 25 Supplement, Vol. to 27, No. 10, Oct. 575 1962, p h65-S TABLE k-11 (Cont'd) L

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0 0 0.2 0.4 0.6 0.8 1.0 1.2 1.4 1.6 1. 8 2. 0 2. 2 2. 4 2.6 2. 8 3.0 3.2 34 Integrated Iseutron Esposure (Evi New). n/ce2 a 10*39 PREDICTED NOTT SHIFT VERSUS REACTOR VESSEL IRRADIATION e 343 Figure 4-12 Supplement-4