ML19322A784

Chapter 4 of Oconee 1,2 & 3 PSAR, Rcs. Includes Revisions 1-6

ML19322A784
Person / Time
Site:	Oconee
Issue date:	12/01/1966
From:	DUKE POWER CO.
To:
References
NUDOCS 7911250003
Download: ML19322A784 (54)

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TABLE OF CONTEITIS Section Page 4

REACTOR COOIANT SYSTEM 4-1 k.1 DESIGN BASES 4-1 4.1.1 PERFORMANCE OBJECTIVES 4-1 4.1.2 DESIGN CHARACTERISTICS 4-1 4.1.2.1 Design Pressure 4-1 4.1.2.2 Design Temperature 4-2 4.1.2 3 Reaction Ioads 4-2 k.1.2.4 Seismic Icads 4-2 4.1.25 Cyclic Icads 4-2 4.1.2.6 Water Chemistry 4-2 4.1 3 EXPECTED OPERATING CONDITIONS 4-2 4.1.4 SERVICE LIFE 4-3 4.1.4.1 Material Radiation Damage 4-3 4.1.4.2 Unit Operational Thermal Cycles 4-3 4.1.4 3 Operating Procedures 4-4 4.1.4.4 Quality Manufacture 4-5 4.1 5 CODES AND CIASSIFICATICNS 4-6 4.2 SYSTEM DECRIPTION AND OPERATION 4-6 4.2.1 GENERAL DESCRIPTION 4-6 4.2.2 MAJOR COMPONENTS 4-6 4.2.2.1 Reactor Vessel 4-6 4.2.2.2 Pressurlzer 4-7 4.2.2.3 Steam Generator 4-8 4.2.2.4 Reactor Coolant Pumps 4-10 4.2.25 Piping 4-11 l

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O COIESES (Cont'd)

V Section Page 4.2 3 PRESSURE-RELIEVING DE7 ICES 4-11 4.2.4 ENVIROIiMEEAL PROTECTION 4-11 k.2 5 MATERIAIS OF CONSTRUCTION 4-12 4.2.6 MAXDUM HEATDIG AND COOLING RATES 4-15 4.2 7 LEAK DEFECTION 4-15 43 SYSTEM DESIGN EVAIUATION 4-16 431 SAFEIY FACTORS 4-16 4 3 1.1 Pressure Vessel Safety 4-16 4.3 1.2 Piping 4-21 4313 Steam Generator 4-22 4.3 2 RELTANCE ON DEERCOMTECTED SYSTDIS 4-23 433 SYSTEM DEEGRITY 4-23 434 PRESURE RELIEF 4-23 435 REDUNDANCY 4-24 4 3.6 SAFEIY ANALYSIS 4-24 4.3 7 OPEFATIONAL LIMITS 4-24 4.4 TESTS AND DISPECTIONS 4-25 4.4.1 COMPONDE IN-SERVICE INSPECTION 4-25 4.k.2 REACTOR SYSTD4 TESTS AND DISPECTIONS 4-25 4.4.2.1 Reactor Coolant System Precritical and Hot Leak Test 4-26

4. 4.2.2 Pressurizing System Precritical Operational Test 4-26 4.4.2 3 Pressurizer Surge Line Temperature Gradient Test 4.26

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4.4.2.4 Relief System Test 4-26 00 00300 1

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j CONTENTS (Cont'd)

Section Pa6e 4.4.2 5 Unit Power Startup Test 4-26 4.4.2.6 Unit Power Heat Balance 4-26 4.4.2 7 Unit Power Shutdown Test 4-26 4.4 3 MATERIAL IRBADIATION SURVEILIANCE 4-26

4.5 REFERENCES

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LIST OF TABIES (At rear of Section)

Table No.

Title Pa6e 4-1 Tabulation of Reactor Coolant System Pressure Settings 4-291 4-2 Reactor Vessel Design Data 4-29 4-3 Pressurizer Design Data 4-30 4-4 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 Codes and Classifications 4-33 4-10 Materials of Construction 4-34 4-11 References for Figure 4 Increase in Transition Temperature Due to Irradiation Effects for A302B Steel 4-35 Ov 00 00302 4-iv

LIST OF FIGURES (At rear of Section)

Figure No.

Title 4-1 Reactor Coolant System 4-2 Reactor coolant System Arrangement Elevation 4-3 Reactor Coolant System Arrangement Plan 4-4 Nil-Ductility Transition Temperature Increase Versus Integrated Neutron Exposure for A302B Steel 4-5 Reactor Vessel 4-6 Pressurizer 4-7 Steam Generator 4-8 Steam Generator Heating Regions 4-9 Steam Generator Heating Surface and Downcomer Ievel Versus Power 4-10 Steam Generator Temperatures 4-11 Typical Reactor Coolant Pump 4-12 Predicted NDTT Shift Versus Reactor Vessel Irradiation i

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PZACTOR C00IAUT SYST31 V

h.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 major coolant system components is shown in Figure 4-1.

The cool-ant system physical arrangement is shown in Figures 4-2 and 4-3 The reactor coolant system is designed to meet the following codes:

Piping and Valves - ASAB31.1-1955 (Pressure Piping) including nuclear cases.

Punp Casing - ASE Boiler and Pressure Vessel Code,Section III, Nuclear Vessels.

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

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

Reactor Vesse1 - ASE Boiler and Pressure Vessel Code,Section III, Nuclear Vessels.

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Welding Qua11fications - ASE Boiler and Pressure Vessel Code, v

Section IX.

To assist in the review of the system drawings, a standard set of symbols and abbreviations have been used and are presented in summary in Figure 9-1.

h.1.1 PERF0lMuiCE OBJECTIVES The reactor coolant system is designed to contain and circulate reactor cool-ant at pressures and flows necessary to transfer the heat generated in the re-actor 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 mod-erator and reflector, and as a solvent for the soluble poison (boric acid) utilized in chemical shim reactivity control.

As the coolant energy and radioactive :.ateria1 container, 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 which escape the primary barrier, the core cladding.

h.1.2 DESIGN CHARACTERISTICS 4.1.2.1 Design Pressure The reactor coolant system design, operating, and control set point pressures l, ' "

are listed in Table 4-1.

The design pressure allows for operating transient pressure changes. The selected design margin considers core thermal lag, e

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coolant transport times and pressure drops, instrumentation and control re-sponse characteristics, and system relief valve characteristics. The design pressures 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 antic-ipated coolant temperature in that component under all nor::n1 and transient load conditions. The design and operating temperatures of the respective sys-tem components are listed in Tables 4-2 thrcugh 4-6.

4.1.2 3 Reaction Icads All components in the reactor coolant system are supported and interconnected so that piping reaction forces result in combined mechanical and thermal stresses in equipment nozzles and structural valls within established code limits. Equip-ment and pipe supports are designed to absorb piping rupture reaction loads for elimination of secondary accident effects such as pipe motion and equipment foundation shifting.

1.1.2.4 Seismic Icads 3

Reactor coolant system components are designated as Class I equipment, 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 will be designed in accordance with Appendix 5A.

4.1.25 Cyclic Icads All components in the reactor coolant cystem are desi ned to withstand the 6

effects 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 chown in Table 4-7 During unit startup and shutdown, the rates of temperature and pressure changes are limited.

4.1.2.6 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 EXmnsu OPERATING CONDITIONS Throughout the load range from 15 to 100 per cent power, the reactor coolant system is operated at a constant average te=perature.

Reactor coolant system pressure is controlled to provide sufficient overpressure to maintain adequate core subcocling.

The mininnm operating pressure is established from core thermal analysis. This analysis is based upon the maximum expected inlet and outlet temperatures, the I

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maximum reactor power, the minimum DNBR required (including instrumentatinu V

errors and the reactor control system deadband), and a core flow distributio factor. The maximum operating pressure is established on the basis of ASME Code relief valve characteristics and the margins required for normal pressure variations in the system.

Pressure control between the preset maximum and minimum limits is obtained directly by pressurizer spray action to suppress high pressure and pressurizer heater action to compensate for low pressure.

Normal operational lifetime transient cycles are discussed in detail in 4.1.4.

4.1.4 SERVICE IlFE The service life of reactor coolant system pressure components depends upon i

the end-of-life material radiation damage, Unit operational thermal cycles, quality manufacturing standards, environmental protection, and adherence to i

established operating procedures.

In the following discussion each of these life-dependent factors will be discussed with regard to the affected compo-nents.

4.1.4.1 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 sub, ject to material radiation damage. To assess the potential radiation damage at the end-of-reactor service life, the maximum exposure from fast neutrons b

(E>1.0mev)hasbeencomputedtobe30x1o19n/cm2 over a 40 year life with V

an 80 per cent load factor. Reactor vessel irradiation exposure calculations are described in 3 2.2.17 For this neutron exposure, the predicted Nil-Ductility Transitign Temperature (NDTT) shift is 250 F based on the curve shown in Figure 4-4.ll) 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 4, represents irradiated material test results and was compiled from the reference documents listed in Table 4-11.

To evaluate the NUIT 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 (4.4.3).

4.1.4.2 Unit Operationt.1 Thermal Cycles To establish the service life of the reactor coolant system components as re-quired by the ASE III, for Class "A" vessels, the unit operating conditions which 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 4-7 under the title " Design Cycles". The estimated actual cycles based on a review of existing nuclear stations operations are also pro-vided in Table 4-7 Table 4-9 lists those components designed to ASE III -

Class "A".

The effect of individual transients, and the sum of these tran-sients, are evaluated to determine the fatigue usage factor during the detail

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design and stress analysis effort. As specified in ASME III Paragraph 415 2 (d)(6), the cumulative fatigue usage factor will be less than 1.0 for the de-sign cycles listed in Table 4-7 The transient cycles listed in Table 4-7 are conservative and complete in that they include all significant modes of normal and emer6ency operation. The es-timated frequency basis for the design transient cycles are listed in Table 4-8.

A heatup and cooldown rate of 100 F/ hour will be used in the analysis of Tran-sients 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 may bc required. Subsequent to a nor-mal refueling operation only the reactor vessel closure seals are hydrotested for pressare integrity; therefore, reactor coolant system hydrotests prior to startup are not included.

h.1.4 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 temperature operation is performed in accordg ge with a

" fracture analysis diagram" as published by Pellini and Puzak.W)

At temperatures below the nil-ductility transition temperature (IDIT) and the desi n transition temperature (DPI), which is equal to IDIT + 60 F, the pres-5 sure vessels will be operated such that the stress levels will be restricted to a value which will prevent brittle failure. These icvels are:

a.

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

b.

From the temperature of D'IT minus 200 F to DTT, a maximum stress which will 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 will not c ur.

This statement is ba<

on data which has been reported by Robertson 3 and Kihara and Masubichi in published literature It can be shown that stress limits can be controlled by imposing operating ( pocedures which control pressure and temperature during heatup and cooldown.

) This procedure will insure that the nominal stress levels do not exceed those as specified in a through c above.

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4.1.4.4 Quality Manufacture Material selection is discussed in detail in 4.2 5 After receipt of the material a program of qualification of all welding and heat treating processes which could affect mechanical or metallurgical prop-erties of the material during fabrication is undertaken. The purpose of this program is to establish the properties of the material as received, and to certify that the mechanical properties 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 same as used for production.

c.

Subjecting qualification test plates to destructive tests to establish the following:

(1) Tensile strength, i

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(2) Ductility.

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

After completion of the qualification test program, production welding 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 finished vessels.

In-process and final dimensional inspections are made to insure that parts and assemblies meet the drawing requirements, and an "as-built" record is kept of these dimensions for future reference.

All welders are qualified or requalified as necessary in accordance with The Babcock & Wilcox Company and ASME IX requirements. Each lot of welding elec-I trodes and fluxes is tested and qualified before release to insure that re-quired mechanical properties and as-deposited chemical properties can be met.

Electrodes are identified and issued only on an approved request to insure that the correct materials are used in each weld. All welding electrodes and fluxes are maintained dry and free from contamination prior to use.

Records are main-tained and reviewed by welding engineers to insure that approved procedures and materials are being used.

Records are maintained for each weld joint and in-clude the velder's name, essential weld parameters, and electrode heat or lot D

number.

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

a.

Radiography, including x-ray, high voltage linear accelerator, or radioactive sources, will be used as applicable to detemine the acceptability of pressure integrity welds and other welds as spec-ifications require.

b.

Ultrasonics is used to examine all pressure-integrity raw material, the bond between corrosion-resistant cladding to base material, and pressure-containing welds.

c.

Magnetic Particle hn

• ntion is used to detect surface or near sur-face defects in machined weld grooves prior to velding, ccepleted weld surfaces, and the co=plete external surface of the vessels in-cluding veld seams after final heat treatment.

d.

Liquid Penetrant is used to detect surface defects in the weld deposit cladding, nonmagnetic materials, and closure studs.

The completed reactor vessel assembly will be shipped as a unit from the fab-rication shop to the station site. The completed reactor closure head will be shipped in like manner.

4.1.5 CODES AND C1ASSIFICATIONS All pressure-containing components of the reactor coolant system are designed, fabricated, inspected, and tested to applicable codes as listed in Table 4-9 4.2 SYSTD4 DESCRIPTION AND OPERATION 4.2.1 GENERAL DESCRIPTION The reactor coJ. ant system consists of the reactor vessel, two vertical once-through steam generators, four shaft-scaled coolant circulating pumps, an electrically heated pressurizer, and interconnecting pipin6 The system is arranged as two heat transport loops, each with two circulating pu=ps and one steam generator. Reactor system desi$n data are listed in Tables 4-2 throu6h 4-6, and a system schematic is shown in Figure 4-1.

Elevation and plan views j

of the arrangement of the major components are shown in Figures 4-2 and 4-3 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 flange to which a removable ree.: tor closure head is bolted. The reactor closure head is a spherically dished head welded to a ring flange.

Tne reactor vessel has six major nozzles for reactor coolant flow, 69 control rod drive assembly nozzles mounted on the reactor vessel head, and two emer-gency injection system nozzles--all located above the core. The vessel closure seal is fomed by two concentric 0-rings with provisions between them for leak

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The reactor vessel, nozzle design, and seals incorporate the ex-tensive design and fabrication experience accumulated by B&W.

Fifty-one in-core instrumentation nozzles are located on the lower head.

The reactor closure head and the reactor vessel lange are joined by sixty 6-1/2in,diameterstuds. Two metallic 0-rings seal the reactor vessel when the reactor 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 supported on lugs welded to the outside of the vessel.

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 assemblies, control cluster assemblies, surveillance specimens, and incore in-strumentation.

The reactor vessel shell material is protected against fast neutron flux and gama 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 4.3 1.

Oh Stop blocks welded to the reactor vessel inside wall limit reactor internals andcoreverticaldropto1/2in.orless,andpreventrotationaboutthever-tical axis in the unlikely event of a major internals component failure.

Surveillance specimens made from reactor steel are located between the reactor vessel wall and the thermal shield. These specimens will be examined at se-1ected intervals to evaluate reactor vessel caterial NITIT changes as described in 4.4 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 Figures 3 46 and 3 47.

Reactor vessel design data are listed in Table 4-2.

4.2.2.2 Pressurizer e

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 pressurizer establishes and maintains the reactor coolant pressure within prescribed limits and provides a surge chamber and a water i

reserve to accommodate reactor coolant volume changes during operation.

The pressurizer is a vertical cylindrical vessel connected to the reactor out-let piping by the surge line. The pressurizer vessel is protected from thennal shock by a thermal sleeve on the surge line and by a distribution baffle located above the surge line entrance to the vessel.

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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 ca-pacity. The capacity of these valves is discussed in 4.3 4.

The relief valves discharge to a quench tank located within the reactor building. The quench 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.

The pressurizer contains replaceable electric heaters in its lower section and a water spray nozzle in its upper section to maintain the stess and water at the saturation temperature corresponding to the desired reactor cooh.nt system pres-sure. During outsurges, as the pressure in the reactor decreases, some of the water in the pressurizer flashes to steam to mintain pressure. Electric heaters are actuated to restore the norm 1 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.

h.2.2 3 Steam Generator The general arrangement of the steam generators is shown in Figure 4-7, and de-sign data are tabulated in Table 4-4.

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 throu6h the tubes, and steam is generated on the shell side. The high pressure parts of the unit are the hemispherical heads, the tubesheets, and the straight Inconel(*) tubes between the tubesheets. Tube sup-ports hold the tubes in a uniform pattern along their length.

The shell, the outside of the tubes, and the tubesheets form the boundnries of the steam-producing section of the vessel. Within the shell, the tube bundle is surrounded by a shroud, which is in two overhpping sections with the upper section the larger of the two in diameter. The upper part of the annulus be-tween the shell and baffle is the superheater outlet, while the lower part is the feedwater inlet-heatin6 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 bottcm heads, m n-vays on both heads, and a drain nozzle for the bottom head. Venting of the re-actor coolant side of the unit is accomplished by a vent connection on the re-actor cochnt inlet pipe to each unit. The unit is supported by a skirt attached to the bottom head.

Reactor coolant water enters the steam Benerator at the upper plenum, flows down the Inconel tubes while transferring heat to the secondary shell-side fluid, and

• Inconel is a trade name of an alloy mnufactured by the International Nickel Company.

It also has substantial co= mon usage as a generic description of a Ni-Fe-Cr alloy confoming to ASTM Specification SB-163 It is in the latter context that reference is mde here.

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exits through the lower plenum. Figure 4-8 shows the flow paths and steam generator heating regions.

Four heat transfer regions exist in the steam generator as feedwater is con-verted to superheated steam. Starting with the feedvater inlet these are:

a.

Feedwater Heating Feedwater is heated to saturation temperature by direct contact heat exchange. The feedwater entering the unit is sprayed into a feed heating annulus (downcomer) formed by the shell and the baffle around the tube bundle. The steam which heats the feedwater to saturation is drawn into the downcomer by condensing action of the relatively cold feedwater.

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 fonced 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. Consequently, 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-creases almost uniformly until DIB, i,e., departure of nucleate boil-ing, 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, hest flux, and mass velocity.

c.

Film Boiling and Superheated Steam Dry saturated steam is produced in the film boiling region at the upper end of the tube bundle. Baffles are used in this region to obtain higher vapor velocity and crossflow for efficient heat trans-

fer, d.

Superheated Steam Saturated steam is raised to final temperature in the baffled super-heater 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 boiling p/

section is proportional to loaa The surface available for super-y heating varies inversely with load, i.e., as load decreases the super-n heat section gains from the nucleate and film boiling regions,

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Mass inventory in the steam generstor increases with load as the len6th of the heat transfer rc6 ons varies.

1 The simple concept with ideal counterflow conditions results in hi6hly stable flow characteristics on both the reactor coolant and se condary sides. The hot reactor coolant fluid is cooled uniformly as it flows downward. The secondary side = ass flow is low, and the majority of the pressure drop is due to the static effect of the mixture. The boiling in the steam generator is some-what similar to " pool boiling", except that there is mtion upward that per-mits some parallel flow of water and steam.

A plot of reactor coolant and steam temperatures versus power is shown in Fig-ure 7-5 As shown, both steam pressure and average reactor coolant temperature are held constant over the load ran8e from 15 to 100 per cent full power. Con-stant steam pressure is obtained by a variable two-phase boiling length (see F1 ure 4-8), and by the regulation of feed flow to obtain proper steam generator 6

secondary mass inventory.

In addition to average reactor coolant temperature, reactor coolant flow is also held constant. The difference between reactor coolant. inlet and outlet temperatures increases proportionately as load is in-creased. Saturation pressure and temperature are held constant, thereby re-sulting in a variable superheater outlet temperature.

Figure h-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 des 1 n can be used.

6 Control of the shell temperature is achieved by the use of direct contact steam that heats the feedwater 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 tem-perature differentials between tubes and shell.

A discussion of the B&W once-through steam generator development program is presented in Appendix hA.

The steam generator design and stress analysis will be performed in accordance with the requirements of the ASME III as described in 4 31.1.

4.2.2. 4 Reactor Coolant Pu=ps The general arrangement of a reactor coolant pump is shown 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 suction and hori-I zontal discharge. Each pump has a separate, single-speed, water-jacketed, top-mounted motor, which is connected to the pump by a shaf t coupling.

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Shaft sealing is accomplished in the upper part of the pump housing using a

'C throttle bushin6, a seal chamber, a mechanical seal, and a drain chn=ber in series. Scal 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 nomal y operates at a pressure of approximately 50 psig and a temperature of 05 to 100 F.

However, it is designed for full reactor cool-ant system pressure and, if seal chamber coolin6 were maintained, would con-tinue to operate satisfactorily without seal water injection for several weeks.

The outboard drain chamber weald further prevent leakage to the reactor build-ing if deteriomtion of the mechanical 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-actin 6, oil-lu-bricated thrust bearing are located in the pump motor. The 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. Oil pressure required for bearing lubrication is maintained by in-ternal pumping provisions in the motor, or by an external system if required for " hydraulic-jacking" of the bearing surfaces for startup.

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 substan-

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tiate that they conform to the initial test pump characteristics.

4.2.2 5 Piping The general arrangement of the reactor coolant system piping is shown in Figures 4-2 and 4-3 Piping design data are presented in Table 4-6.

In addition to the pressurizer surge line connection, the piping is equipped with welded connections for pressure taps, temperature elements, vents, drains, decay heat removal, and emergency high pressure injection water. Thermal sleeves are provided in the pressurizer surge line and the emergency high pressure injection line connections.

4.23 PRESSURE-RELIEVING 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 valve effluent. The schematic arrangement of the relief devices is shoEn in Figure 4-1.

Since all sources of heat in the system, i.e., core, pressurizer heaters, and reactor coolant pumps, are interconnected 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 p

The recctor coolant system is surrounded by concrete shield walls. These walls

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. (

provide shielding to permit access into the reactor building for inspection and maintenance of miscellaneous rotating equipment during full power operation and H

c ' 10 f 00 00314 4-11

for periodic calibration of incore detectors. These shielding walls act as missilt protection for the reactor building liner plate.

Lateral bracing will be provided near the steam generator upper tube sheet elevation 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.

Concrete slabs over the reactor coolant system are also provided for shielding and missile dama6e protection.

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 h-10.

All reactor coolant system caterials exposed to the coolant are corrosion-resistant materials consisting of 304 or 316 ss, weld deposit 304 ss cladding, Inconel (Ni-Cr-Fe), and 17 4 PH (H1100). These materials were chosen for spe-cific purposes at varfous locations within the system because of their superior compatibility with the reactor coolant.

Periodic analyses of the coolant chemical composition will be performed to mon-itor the adherence of the system to the reactor coolant water quality listed in Table o-4.

Maintensuce of the water quality to minimice corrosion is per-formed by the chemical addition and sampling system which is described in de-tail in 9 2.

The feedwater quality entering the steam generator will be held within the limits listed in Table 9-3 to prevent deposits and corrosion inside the steam generators. This required feedwater quality has been successfully used in comparable once-throu6h, nonnuclear steam generators. The phenomena of stress-corrosica crackin6 and corrosion fati6ue are not generally encountered unless a combination of elements in varying degrees is present. The necessary condi-tions are a susceptible alloy, an aggressive environment, a stress, and time.

It is characteristic of stress corrosion that combinations of alloy and environ-ment that result in cracking are usually quite specific. Environments that have shown to cause stress-corrosion cracking of stainless steels are free alkalinity in the presence of a concentrating mechanism and the presence of chlorides and free oxy 6en. With regard to the former, experience has shown that chemical

" hideout" or deposition of chemicals on the surface of tubes can occur in a

" steam blanketed" area within a steam generator. In the presence of this envi-ronment, stress-corrosion cracking can occur in stainless steels havin6 the nominal residual stresses resulting from normal manufacturing procedures. The once-through steam generator contains Inconel tubes. Testing to investigate the susceptibility of heat exchanger construction materials to stress corrosion in caustic and chloride aqueous solutions indicated that Inconel Alloy 600 has excellent resistance to general and pitting-type corrosion in severe operating water conditions. Extensive operating experience with Inconel units has con-firmed tnis conclusion.

Inconel steam generator tubing is being used in the c

B 's uu s 00 00315 h-12

.b German Nuclear Ship, " Otto Hahn", and the tubing has been tested under service temperature and pressure conditions and in a high 02 vater environment as fol-lows:

Water Temperature 572 - 650 F Pressure 2,000 psig pH 69 0.040 ppm 02

)

l Tubing inspected after 5,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> showed no signs of stress-corrosion crack-ing or other detrimental effects even though the 02 was 0.040 ppm and consid-

)

erably above the maximum that will be used in the Oconee Nuclear Station steam generators (Tables 9-3 and 9-4).

This reaffirms the conclusion that Inconel is a satisfactory material for this service.

All external insulation of reactor coolant system components will be compatible with the component materials.

The reactor vessel is insulated with :etallic reflective insulation on the cylindrical shell exterior. The closure flanges and the top and bottom heads in the area of corrosion-resistant penetrations will be insulated with low halide-content insulating material. All other ex-ternal corrosion-resistant surfaces

..n the reactor coolant system will be insu-lated with lov or halide-free insule. ting material as required.

Os The reactor vessel plate material,pp(. site the core is purchased to a specified Charpy V-notch test result of 30

-lb or greater at a corresponding nil-ductility transition temperature t 'DTT) of 10 F or less, and the material will be tested to verify conformity to :pecified requirements and to determine the actual NDTT value.

In addition, nis plate will be 100 per cent volumetrically inspected by ultrasonic test using both normal and shear wave.

The remaining material in the reactor vessel, and other rea: tor coolant sys-tem components, is purchased to the appropriate design code requirements and specific component function.

The reactor vessel material is heat-treated specifically to obtain good notch-ductility which will ensure a lov NDTT, and thereby to give assurance that the finished vessel can be initially hydrostatically tested and operated at room temperature without restrictions. The stress limits established for the reac-tor vessel are dependent upon the temperature at which the stresses are ap-plied. As a result of fast neutron absorption in the region of the core, the material ductility will change. The effect is an increase in the NDTT.

The l

predicted end-of-life NDTT value of the reactor vessel opposite the core is 260 F or less. The predicted neutron exposure and NDTT shift are discussed in l

4.1.4.

The unirradiated or initial NDTT of pressure vessel base plate material is pres-ently measured by two methods. These methods are the drop weight test per ASTM E208 and the Charpy V-notch impact test (Type A) per ASTM E23 The NDTT is de-l fined in ASLM E208 as "the temperature at which a specimen is broken in a series b3 of tests in which duplicate no break performance occurs at a 10 F higher tem-perature". Using the Charpy V-notch test, the NDTT is defined as the temperature c

1 ROU of:

00 00316 4-35

at which the energy required to break the specimen is a certain " fixed" value.

For SA 302B steel the ASME: III Table N-332 specifies an energy value of 30 ft-lb.

This value is based on a correlation with the drop weight test and will 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 ICIT.

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

The =aterial for these tests will be treated by the methods as outlined by ASE III Paragraph U-313 The test coupons will 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 performed by the material sup-plier to certify the material as delivered to B&W. The exact test coupon lo-cations are reviewed and approved by B&W to insure compliance with the applicable ASE Code and specifications.

In accordance with ASE III Paragraphs N-712 and N-713, B&w performs Charpy V-notch impact tests on heat-affected zone (HAZ),

base metal, and weld metal on all pressure vessel test plates.

Differences of 20 to 40 F in ICIT have been observed between T/4 and the sur-face in heavy plates. The T/4 location for Charpy V-notch impact specimens is conservative since the NDTI' of the surface material is lower than that of the internal material.

The reactor vessel design includes curveillance specimens which will permit an evaluation of the neutron exposure-induced shift on the material nil-ductility transition temperature properties.

The material irradiation surveillance program is described in 4.4 3 O

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00 00317 J U no 4-14

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4.2.6 MAXIIEi EATING AND COOLING RATES The normal 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.

It is anticipated that the reactor vessel per=issible heatup rate can exceed 100F/hr. Consequently, there are sufficient electrical heaters installed in the pressurizer to permit a heatup rate, starting with minimum vater level, of 150F/hr. The exact final rates are determined during the detail design and stress analysis of the vessel.

The fastest cooldown rates which result from the break of a =ain steam line are discussed in 14.1.2 9.,

4.2 7 LEAK DET ETION To minimize leakage from the reactor coolant system all components are inter-connected by an all-welded piping system.

Some of the components have access openings of a flanged-gasketed design. The largest of these is the reactor vessel closure, which has a double metal 0-ring seal with provisions for mon-itoring for leakage between the 0-rings. Other openings and appurtenances to the reactor coolant system which are possible sources of leakage are tabulated in detail along with the maximum expected rates of leakage in Section 11.

With regard to the reactor vessel, the probability of a leak occurrin6 is con-sidered to be remote on the basis of reactor vessel design, fabrication, test, inspection, and operation at temperatures above the material NDTI as described in 4 3 1.

Reactor vessel closure leakage vill be zero from the annulus be-tween the metallic 0-ring seals during vessel steady-state and virtually all transient operating conditions.

Only in the event of a rapid transient opera-tion, such as an emergency cooldown, would there be some leakage past the in-nermost Oering seal. A stress analysis on a similar vessel design indicates this leak rate would be approximately 10 cc/ min through the seal monitorinr taps to a drain, and no leakage vill occur past the outer 0-ring seal. TIL exact nature of this transient condition, and the resulting small leak rate, will be determined by a detailed stress analysis.

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

a.

Instru=entation in the control room will indicate the addition rate of makeup water required to =aintain normal vater level in the pressurizer.

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 room instrumentation vill indicate the existence of a change in Q

the water level in the reactor building sump.

00 00318 a

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If any one of the above methods indicates an excessive reactor coolant leakage rate during operation, the reactor vill be taken to a cold shutdown, and the cause of the problem vill be deter =ined.

43 SYSTEM DESIGN ~ EVALUATION 431 SAF E Y 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 psic at a nominal te=perature 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 4-9 Aside from the safety factors introduced by code require =ents and quality con-trol programs, as described in the following paragraphs, the reactor coolant system functional safety factors are discussed in Sections., and 14.

4 3 1.1 Pressure vessel Safety The safety cf the nuclear reactor vessel and all other reactor coolant system pressure vessels is dependent upon four major factors:

(1) design and stress analysis, (2) material selection and fabrication, (3) quality control, and (4) proper operation. The special care and detail used in i=plementing these factors in pressure vessel manufacture are briefly described as follows:

4 3 1.1.1 Design and Stress Analysis These pressure vessels are des 16ned to the requirements of the ASME III. This code is a result of ten years of effort by representatives from industry and government who are skilled in the design and fabrication of pressure vessels.

~

It is a comprehensive code based on the most applicable stress theory.

It re-quires a stress analysis of the entire vessel under both steady-state and tran-sient operations. The result is a complete evaluation of both primary and sec-ondary 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 from Table 4-7, the fatigue evaluation 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 & IJilcox Company has confin::ed 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 agreement with

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00 00319 4-16

O-those predicted by the theory. B8M 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 conditions. A detail design and analysis of every part of the vessel is prepared as follows:

a.

The vgssel size and configuration are set to meet the process require-ments, the thickness requirements due to pressure and other 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 yressure loading and thermal stresses are calculated.

Computer programs are used in this devel-opment.

c.

An evaluation of the stresses through the vessel is performed using as a 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 ves-sel), membrane plus primary bending stress (to insure a distortion-free vessel), secondary stress (to insure a vessel that will not pro-s S

gressively deform under cyclic loading), and peak stresses (to insure a vessel of maximum fatigue life).

Much of the stress analysis mentioned in the above listing is statically inde-terminate. Hence, when an evaluation of these stresses, as mentioned in e above, shows them to be in excess of permissible ASME III values, corrective design changes are made and the procedure reiterated.

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 basis, loading conditions, etc., and will sum-marize the conclusions to permit independent checking by interested parties.

4.3 1.1.2 Quality Control In-process and final dimensional inspections are made to 1nsure that parts and assemblies meet the drawing requirements, and au "as-built" record is kept of

)

these dimensions for reference. A temperature-controlled gage-room is maintained to keep all measuring equipment in proper calibration, and personnel supervising this work are trained in formal programs sponsored by gace equipment manufacturers.

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00 00320 1

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O The practice of applied radiography is being continually improved to enhance flav 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.

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

c.

An 0.080 in. lead filter is used at the film to absorb "broadbeam scatter", when using high voltace equipment (above 1 mev).

d.

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

e.

Densities of radiographs are controlled by the use of densitometers, f.

Double film technique is used on all genma-ray exposures as well as high voltage exposures.

g.

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

Energies used are controlled to be in the optimum ran;e.

Ultrasonics is one of the most useful inspection tools.

It is being used as follows:

a.

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

b.

In order to detect laminations which are normally parallel to the sur-face, plates are also inspected by a shear wavel The bond between cladding and base material is inspected by ultra-c.

sonics.

d.

All plate is 100 per cent volumetrically inspected by ultrasonics usin; both nomal and shear wave.

e.

Personnel conducting ultrasonic inspections are given extensive training.

The magnetic 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 magnetic particle method after removal of backup strips.

e) - q, pp 00 00321 h-18 i

A b.

Veld preparations are inspected by the =agnetic particle method.

b c.

A complete external surface inspection of the entire vessel includ-ing veld seams is performed after all heat treatment.

d.

Personnel usin3 this =ethod are trained by B&W in addition :o attend-ing formal procrsms conducted by the equipment manufacturers.

The J.1 quid penetrant examination is used to aid in detecting surface defects and is particularly adaptable to the nonmagnetic =aterials such as stainless steel.

It is presently being used as follows:

a.

Inspection of veld-deposited cladding.

b.

Inspection of reactor vessel studs.

c.

Personnel usin3 this method of examination are trained by B&W in addition to attending formal programs conducted by the equipment manufacturers.

The prh2ary purpose of the above quality control procedures and methods is to loca+,e, define, and determine the size of material defects to allow an evalua-tion of defect acceptance, rejection, or repair.

The site of defect that can conceivably contribute to the rupture of a vessel

.(N depends not only on the size effect, but also on the orientation of the defect, the magnitude of the stress field, and temperature. The correlation of these

.s major parameters has been done by Pellini and Puzak(21 vho have prepared a

" fracture analysis diagram" which is the basis of vessel operation from cold startup and shutdown to full pressure and temperature operation.

1 The diagram predicts that, for a given level of stress, larger flav sizes vill be required for fracture initiation above the NDIT temperature. For example, atstressesintheorderof3/4yieldstrength,aflawintheorderof8to10 in. may be sufficient to initiate fracture at temperatures below the UDIT te=-

perature. However, at UDIT + 30 F, a flav of 1-1/2 times this size may be re-i quired for initiation of fracture. While at a temperature of ITDIT + 60 F, brit-tle fracture is not possible under elastic stresses because brittle fracture propagation does not take place at this te=perature. Fractures above this tem-perature are of the predominantly ductile type, and are dependent upon the mem-ber net section area and section modulus as they establish the applied stress.

Stud forgings vill be inspected for flaws by two ultrasonic inspections. An axial longitudinal beam inspection vill be performed. The rejection standard villbeloss-of-back-reflectionSreaterthanthatfroma1/2in.diameterflat bottom hole. A radial inspection vill be made using the longitudinal beam tech-nique. This inspection vill carry the same rejection standards as the axial inspection.

In addition to ultrasonic test, liquid penetrant inspection vill be performed on the finished studs.

The stress analysis of the studs vill include a fatigue evaluation.

It is not O

expected that fatigue evaluation vill yield a significantly high usage factor d

for the 40 year design life. Therefore, there vill be no planned frequency for stud replacement.

If an indication is found when the studs are inspected a

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T during refuelin3, as described belov, the stud vill be replaced.

One-third of the studs vill be visually examined and dye-penetrant inspected at every refueling. Any positive indications found v111 be cause for rejection.

The reactor vessel closure contains sixty 6-1/2 in diameter studs.

The stud =aterial is A-540, Grade 323 (ASIS III, case 1335) which has a mini-mum yield strength of 130,000 psi. The studs, when tightened for operating conditions, vill have a tensile stress of approximately 30,000 psi. This means:

At Operatin: Conditions (2185 psig):

a.

10 adjacent studs can fail before a leak occurs.

b.

25 adjacent studs can fail before the recaining studs reach yield

strength, 26 adjacent studs can fail before the remaining studs reach the ulti-c.

mate tensile strength, d.

43 sy= metrically located studs can fail before the re=aining studs reach yield strength.

4 3 1.1 3 Operation As previously mentioned in 4.1.4, pressure vessel service life is dependent on adherence to established operatin3 procedures. Pressure 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 ASIE III are the bases of designing for fatigue. They are based on fatigue tests of pressure vessels sponsored by the AEC and the Pressure Vessel Research Co=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 abscrption in the region of the core, the reactor vessel material ductility vill chance. The effect is an increase in the nil-ductility transition temperature (NUTI). The determination of the predicted NDIT shift is described in 4.1.4.1.

This NUPI shift is factored into the plant startup and shutdown procedures so that full operating pressure is not attained until the reactor vessel temperature is above the design transition temperature (DIT).

Below the DFr the total stress in the vesso'. vall due to both pressure and the associated heatup and cooldown transient is restricted to 5,000 - 10,000 psi, which is below the threshold 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 followed. Additional information on the determination 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 methods and procedures used in pressure vessel design, not previously

=entioned in 4 3 1.1 above but which are considered conservative and provide an c

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additional margin of safety, follow.

t a.

Use of a stress concentration factor of 4 on assu=ed flaws in calcu-lating stresses.

b.

Use of mini =um specified yield strength of the material instead of the actual values.

c.

Neglecting the increase in yield strength resulting from irradiation effects.

d.

The desi n shift in NUIT as given in 4.1.4.1 is based on maximum pre-6 dicted flux levels at the reactor vessel inside vall surface, whereas the bulk of the reactor vessel material vill experience a lesser expo-sure of radiation and consequently a lower change in NDIT over the life of the vessel.

e.

Because the irradiation dosage is higher at the inside surface of the j

reactor pressure vessel vall, the surveillance specimens will be sub-jected to a greater degree of irradiation, and therefore a larger shift in the NDIT value than vill be experienced by the vessel. The speci-mens lead the vessel with respect to irradiation effects and i=part a degree of conservatism in the evaluation of the capsule specimens.

The material irradiation surveillance program is described in 4.k.3 f.

d Results from the method of neutron flux calculations, as described in 3 2.2.1 7, have increased the flux calculations by a factor of 2 in predicting the nvt in the reactor vessel vall. The conservative assump-tions, uncertainties, and comparisons of calculational codes used in determining this factor c.re discus +ad in detail in 3 2.2.17 The foregoing discussiou presents a detailed description of quality design, fab-rication, inspection, and operating procedures used to insure confidence in the integrity of pressure vessels.

Experience reported by Reference (5), MW, and the satisfactory experience of MW customers support the conclusion that pressure ves-sel rupture is incredible.

h.3 1.2 Piping Total stresses resulting from thermal expansion and pressure, and mechanical and seismic loadings, are all considered in the design of the reactor coolant 1

piping. The total stresses which can be expected in the piping are within the i

maximum code allowables. The pressurizer surge line connection, and the high pressure injection connections, are equipped with thermal sleeves to limit stresses from thermal shock tc. acceptable values. All materials and fabrica-tion procedures will meet the requirements of the specified code. All mate-rial vill be ultrasonically inspected. All velds vill be radiographically in-spected. All interior surfaces of the interconnecting piping are clad with stainless steel to eliminate corrosion problems r.nd to reduce coolant contam-n ination.

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4313 steam cenerator Because the basic concept of the once-through steam generator would indicate the possible existence of differential ther=al expansion-induced stresses in either the tubes or shell, an evaluation of the ther=al loadings has been perfor=ed 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 14.1.2 9 closely simulates the above design premise in a transient man 7:r.

Secondar/' temperature variations during the acci-dent are essentially transient skin effects with the controlling temperature for the tubesheets and tubes being that of the reactor coolant. Thermal stresses for this case vill be belov ASME allowable values. Some tube deformation may occur but this vill be restrained by the tube supports.

During nomal power operation the tubes are hotter than the shell of the steam generator in the amount of 10 F to 20 F depending upon load. The effect is to put the tubes in a slight compression of 3,000 psi at the 20 F maximum temperature difference. This causes no adverse effect on the tubes since this stress is well below the allovable stress of 23,300 psi for sB-163 material (AsME III, case 1336).

Buckling of the tubes does not occur since these are supported laterally at 40 in, intervals along their lengin.

During startup and shutdown operations the tubes are hotter than the shell of the steam generator in the amount of 40 F.

This places the tubes in a compressive stress of 6,000 psi which causes no adverse effect on the tubes since this stress is well below the allowable stress of 23,300 psi for this SB-163 material (ASME III, Case 1336). Buckling of the tubes does not occur since these 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 construct-ed at the same tube size, length, and material as the steem generator, but of seven tubes in number.

It was structurally tasted with a ther=al difference of shell and tube of 80 F for 2,000 cycles. This severe ther=al cycle test was perfomed with a tube-to-shell temperature difference twice as great as the max 1=um expected during startup and shutdown (Transients 1 and 2, Table 4-7). Destructive examination of the unit after this test indicated no adverse effects from fatigue, stress, buck-ling, or tube-to-tubesheet joint leaka6e.

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00 00325 4 22

432 RELTANCE ON INTERCONNECTED SYSTEMS The principal heat removal systems which are interconnected with the reactor coolant system are the steam and feedwater systems and the low pressure in-jection and decay heat removal system. The reactor coolant system is depen-dent upon the.* steam generators, and the steam, feedwater, and condensate sys-tems for decay heat removal from normal operating conditions to a reactor coolant temperature of approximately 250 F.

All vital active components in these systems are duplicated for reliability purposes.

The engineering 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 stored in the feedwater system may be pumped into the steam generators and the resultant steam vented to the i

atmosphere. A turbine-driven, 5 per cent capacity, boiler feed pump will supply water to the steam generators.

i The low pressure injection and decay heat removal system is used to remove decay heat when the thermal driving head of the reactor coolant system is no longer adequate to generate steam. A complete description of this system is presented in 9 5 The heat received by this system is ultimately rejected to the service water system which also contains sufficient redundancy to i

guarantee proper operation. A schematic diagram of the service water system is presented in Figure 9-9 1

433 SYSTEM INTEGRITY Integrity of the reactor coolant system is insured by proper materials selec-tion, fabrication quality control, design, and operation. All components in the reactor coolant system are fabricated from materials initially having a low nil-ductility transition temperature (NDIT) to eliminate the possibility of propagation-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 incorporat-ed into reactor operation to insure operation above NITfT.

The coolant system is designed in accord with ASMS pressure vessel and ASA power piping codes as covered in 4.1.

Relief valves on the precsurizer are j

sized to prevent system pressure from exceeding the design 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 p-ior to initial operation. The largest and most frequently used opening in reactor coolant system, the reactor vessel head, contains provisions for separate hydrostatic pressurization between the 0-ring type gaskets.

i 4.3.4 PRESSURE RELIEF The reactor coolant system is protected against overpressure by safety valves c.

located on the top of the pressurizer.

c 90'0 1 00 00326

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443

The capacity of the pressurizer safety valves is determined from consider-ations of:

(1) the reactor protective system, (2) line pressure 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 dis-charge line, and (4) accident or transient conditions which may potentially cause overpressure.

Preliminary analysis indicates that the hypothetical case of withdrawal of a regulating control cluster assembly bank from a relatively low power provides the basis for establishing pressurizer safety valve capacity. The accident is terminated by high pressure reactor trip with resulting turbine trip.

This accident condition produces a power mismatch between the reactor coolant sys-tem and steam system larger than that caused by a turbine trip without immedi-ate reactor trip, or by a partial load rejection from full load.

The safety valve capacity required to prevent overpressure for this hypotheti-cal condition is approximately 600,000 lb/hr of steam. The safety valves are sized in accordance with the requirements of ASME III.

4.3.5 REDUNDANCY The reactor coolant system contains two steam generators and four reactor coolant pumps. Operation at reduced reactor power is possible with one or more pumps out of service.

For added reliability, power to each pump is normally supplied by one of two electrically separated busses as shown in Figure 8-1.

Two core flooding nozzles are located on the opposite sides of the reactor vessel.

Each nozzle is connected to a core flooding tank and a line from the low pressure injection system. The high pressure injection lines are connected to the reactor coolant system on two of the four coolant inlet pipes, again to further insure water injection.

4.3.6 SAFETY ANALYSIS The components of the reactor coolant system are interconnected by an all-welded piping system.

Since the reactor inlet and outlet nozzles are all located above the core, there is never any danger of the reactor coolant un-covering the core when any other system component is drained for inspection or repair.

4.3.7 OPERATIONAL LIMITS Reactor coolant system heat up and cool down rates are described in detail in 4.1.4 and 4.2.6.

The component stress limitations dictated by material NDTT considerations are described in 4.1.4 and 4.3.1.

The reactor coolant system is designed for 2,500 psig at 650 F.

The normal operating conditions will be 2,185 psig at an average system temperature of 579 F at full power.

In this mode of operation, the reactor vessel outlet temperature is 603 F.

Additional temperature and pressure variations at various power levels are shown on Figure 7-5.

N : 0 (.

00 00527

'~

4-24 (Revised 4-1-67)

Q Reactor trip signals will be fed to the safety and protective system as a re-Q sult of high coolant temperature, high pressure, low pressure, and low flow, i.e., flux-flow comparator. By relating low flow to the reactor power, opera-tion at partial power is feasible with less than four reactor coolant pumps operating. The reactor operating limits are as shown in the following tabula-tion:

Performance Vs Pumps In-Service Reactor Coolant 2 Pumps 2 Pumps Pumps Operating 4 Pumps 3 Pumps (Same Icop)

(2 Icops) 1%ximum Reactor Power, %

100 86 64 60 ReactorCoolantFlow,%

100 74 46 38 Reactoi 14.1.2.6.yrating limits under natural circulation conditions are discussed in 3

Further discussions of the bases for the selection of operational limits are presented in 7 1.2.4 The reactor coolant system is desi5ned 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.

4.4 TESTS AND INSPECTIONS 4.4.1 COMPONENT IN-SERVICE HISPECTION All reactor vessel internals are :emoveable to facilitate vessel inspection should it become necessary.

The pressure shell of the steam generator is completely inspectable on the out-side by removing the thermal insulation. Direct visual examination, and mag-netic particle or ultrasonic techniques, can be used if necessary. The reac-tor coolant side of the heads and tubesheets can be inspected by direct visual techniques by removin6 the 16 in. manway covers in the heads. The inside of the cylindrical portion of the main pressure shell can be inspected through manholes and handholes. The feedwater spray nozzles are removable for in-specting the shell nearby.

The reactor building arrangement provides sufficient space for inspection of 4

the external surfaces of the reactor coolant piping.

4.4.2 REACTOR SYSTEM TESTS AND HISPECTIONS

(

The assembled reactor coolant system will be subjected to the following tests and inspections during final unit construction and initial startup phases.

(

These tests are in addition to the tests in compliance with code requirements.

J 00 00328 1

w w b>

u.

4-2, s 3sJ

4. 4.2.1 Reactor Coolant System Precritical and Hot Icak Test The objective of this test is to de=cnstrats satisfactory preliminary opera-tion of the entire system and its individual components, to check and evaluate operating procedures, and to determine reactor coolant system integrity at nor-cal operating temperature and pressure.

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

4.4.2 3 Pressuriser Surge Line Temperature Gradient Test The temperature at the midpoint of the pressurizer surge line is determined after a period of steady state operation to check temperature gradients.

4.k.2.4 Relief System Test In this test all relief valves are set and adjusted, and operating procedures are evaluated.

k.k.2 5 Unit Power Startup Test This test determines perfor=ance characteristics of the entire unit in short periods of operation at steady state power levels.

k.4.2.6 Unit Power Heat Balance The purpose of this test is to determine the actual reactor heat balance at various power levels to provide the necessary data for calorimetric calibra-tion of the nuclear instrumentation and reactor coolant system flow rate, k.k.2 7 Unit Power Shutdown Test Tne purpose of this test is to check and evaluate the operating procedures used in shutting down the. unit and to determine the overall unit operating characteristics during shutdown operations.

h.k.3 MAT 2 RIAL IRFADIATION SURVEILIANCE Surveillance specimene of the reactor vessel shell section material are in-stalled between the core and inside wall of the vessel shell to monitor the NDTT of the vessel material during operatin6 lifetime.

The type of specimens included in the surveillance program will be Charpy V-notch (Type A) and tensile specimens for measuring the changes in =aterial properties resulting from irradiation. This is in accordance with ASTM 2185, "Reco=r.anded Practice for Surveillance Tests on Structural Materials in Nu-clear Reactors".

O The reactor vessel material surveillance program will utilize a total of eight ecimen capsules. Four capsules will be located close to the inside reactor c

' Q:j t-26 00 00329

(

vessel wall and directly opposite the center portion of the core for each re-

'O actor vessel. The locations of the capsule holder tubes are shown in Figures 3 46 and 3 47 In these positions, the irradiation received by the specimens will be approximately three times that received by the reactor vessel.

The mterial from each reactor vessel will have its initial III71T determined by the Charpy V-notch impact correlation with drop wei6ht tests. The pre-dicted shift or change in the IM of the vessel material resulting from irradiation is discussed in detail in 4.1.4.1 and is shown in Fidure 4-12 relative to years of operation.

The influence of neutron irradiation on the reactor vessel material properties will be evaluated periodically during unit shutdowns for refueling as tabulated below. This will be accomplished by testing samples of the material from each reactor vessel which are contained in the surveillance specimen capsules.

These capsules contain steel coupons from plate, weld, and heat-affected zone material used in fabricating the reactor vessels. Dosimeters are placed with the Charpy V-notch impact specimens and tensile specimens. The dosimeters will permit evaluation of flux as seen by the specimens and vessel wall. 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 ITMT which may be measured in a manner similar to the initial ND1T. These test results can be compared with n

the then-existing data on the effects of neutron flux and spectrum on engineer-1 I

ing materials.

V Schedule for Capsule Removal Exposure Time Unit No.

Capsule IIumber (Yearc from start of unit) 1 1

1 2

5 3

10 4

15 2

5 5

6-15 7

spares 8

Spares The measured neutron flux and NIET may then be compared and evaluated with the initial NIET and predicted NMT shift to monitor the progress of radiation-induced changes in the vessel materials. As the end of reactor design life nears, a significant increase in measured IMT in excess of the predicted NIET shift could be investigated by review of the vessel stress analysis and operating records. If necessary or required in accordance with the advanced p

knowledge available at that time, the vessel transient limitations on pres-g sure and temperature may be altered so that vessel stress limits, as stated in 4.1.4.3 for heatup and cooldown, are not exceeded.

'('

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4-27 00 00330

A }i

{

4.5 REFEE3CES (1) Porse, L., Reactor Vessel Design Considering Endiation Effects, ASIE Paper No. 63-WA-loo.

(2) Pellini, ',l. 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 (3) Robertson, T.

S., Propagation of Brittle Fracture in Steel, Journal of Iron and Steel Institute, Volume 175, Dece=ber 1953 (4) Kihara, H. and Masubichi, K., Effects of Residual Stress on Brittle Frac-ture, Uelding Journal, Volume 38, April 1959 (5) Miller, E. C., The Integrity of Reactor Pressure Vessels, ORNL-I; SIC-15, May 1966.

{J'

'T l

,rs wJ 00003 33 k-28 i

O Table k-1 Tabulation of Reactor Coolant System Pressure Settings l

Item Pressure, psig 1

Design Pressure 2,500 Operating Pressure 2,185 Code Relief Valves 2,500 High Pressure Trip 2,350 High Pressure Alarm 2,300 kw Pressure Alarm 2,150 kw Pressure Trip 2,000 Table 4-2 Reactor Vessel Design Data Item Data Design /Operatin6 Pressure,psig 2,500/2,185 Hydrotest Pressure (cold), psig.

3,125 Design / Operating Temperature, F 650/600 Overall Height of Vessel and Closure i

Head, ft-in.

37 4 l

Straight Shell Thickness, in.

8-7/16 Water Volume, ft) 4,150 Thickness of Insulation, in.

3 Number of Reactor Closure Head Studs 60 ID of Flan 6e, in.

165 ID at Shell, in.

171 Inlet Nozzle ID, in.

28 Outlet Nozzle ID, in.

36 Core Flooding Water Nozzle, in. - /y TABLE h-1, h-2 5

1 4-29

O Table 4-3 Pressurizer Design Data Item Data Design /OperatingPressure,psig 2,500/2,185 Hydrotest Pressure (cold), psig 3,125 Design /OperatingTemperature,F 670/650 Normal Water Volume, ft) 800 3

Normal Steam Volume, ft 700 Surge Line Nozzle Diameter, in.

10 Overall Height, ft-in.

44-0 Table 4-4 Steam Generator Design Data Item Data per Unit DesignPressure,ReactorCoolant/ Steam,psig 2,500/1,050 Hydrotest Pressure (tube side-cold), reactor coolant psig 3,125 DesignTemperature,ReactorCoolant/ Steam,F 650/600 ReactorCoolantFlow,lb/hr 65 66 x 10 Heat Transferred, Btu /hr 4.21 x 109 Steam Conditions at Full Icad, Outlet Nozzles:

6 SteamFlow,lb/hr 5 30 x 10 Steam Temperature, F 570(35Fsuperheat)

Steam Pressure, psig 910 Feedwater Temperature, F 455 Overall Height, ft-in.

73 1/2 Shell OD, in.

147-1/16 3

Reactor Coolant Water Volume, ft 2,030 O

TABLE h-3, h-h 00 00333 u_,,

O Table 4-5 Reactor Coolant Pump Design Data Item Data per Unit Number of Pumps 4

Design Pressure, psig 2,500 Hydrotest Pressure (cold), psig 3,125 Design Temperature, F 650 Operating Speed, rpm 1,180 Pumped Fluid Temperature, F 60 to 580 Developed Head, ft 370 Capacity, gpm 88,000 Hydraulic Efficiency, %

4 ff Seal Water Injection, gym 4 w 65 4 o' To ##

Seal Water Return, gpm 55 'n M

1. 3 T# #T Pump Nozzle ID, in.

28 Overall Unit Height, ft 24 3

Water Volume, ft

. :P3-- Tf Motor Stator Frame Diameter, ft 8

2 Pump-Motor Moment of Inertia, lb-ft 70,000 Motor Data:

Type Squirrel-Cage Induction, Single Speed Voltage b ^^^

f 4av Phase 3

Frequency, cps 60 Starting Across-The-Line Input (hotreactorcoolant),kw

-5,000 C 4##

Input (cold reactor coolant), kw

-7,350 7, V 20 TABLE h-5 00 00334 u_31

Table 4-6 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. 400- /40 Design /OperatingPressure,psig 2,500/2,185 Hydrotest Pressure (cold), psig 3,125 Design /OperatingTemperature,F 650/605 Design /OperatingTemperature (pressuri~zer surge line), F 670/650 Water Volume, ft) 1,900 O

Table 4-7 Transient Cycles Estimated Transient Description Design Cycles Actual Cycles 1.

Heatup, 70 to 579 F and Cooldown, 579 to 70 F 480 80 G.

Heatup, 540 to 579 F and Cooldown, f

579 to 540 F 1,440 770 3

Ramp Ioading and Ramp Unicading 12,000 9,000 4.

Step Ledina Increase 2,000 1,500 5

Step Unioading Decrease 2,000 1,500 6.

Step Ioad Reduction to Auxiliary Icad 160 120 7

Reactor Trip From Full Power 400 300 8.

Miscellaneous Transients 10 5

Above cycles are based on 40 year design life.

TABIE h-6, h-7 q

00 00335

,_32

[

'O Table 4-8 Design Transient Cycles Transient No. (See Table 4-7)

Frequency 1.

Heatup, 70 to 579 F and Cooldown, 579 to 70 F 12 per Year 2.

Heatup, 540 to 579 F and Cooldown, 579 to 540 F 36 per Year 3

Ramp Icading and Ramp Unloading 6 per Week 4.

Step Leading Increase 1 per Week 5

Step bading Decrease 1 per Week 6.

Step Mading Reduction to Auxiliary Load 4 per Year 7

Reactor Trip From Full Power 10 per Year Table 4-9 Reactor Coolant System Codes and Classifications Component Code Classification Reactor Vessel ASME(*)III Class A Steam Generator ASME(* III Class A Pressurizer ASME(* III Class A Reactor Coolant Pump

_ 7:1ut; n i Casing ASME("}III Class A IEEE,(b)

(c) ud ASA(d)

Motor Coolant Piping ASA(')B31.1-1955and Asso-ciated Nuclear Code Cases

(" American Society of Mechanical Engineers, Boiler and Pressure Vessel Code.Section III covers Nuclear Vessels.

(b) Institute of Electrical and Electronics Engineers.

National Electrical Manufacturers Association.

(

American Standards Association No. C30.2-1955 and C50.20-1954.

(e)American Standards Association No. B31.1.

v TABLE h-8, h-9 c

u_33 00 00336

O Table 4-10 Materials of Construction Component 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 Thermal Shield and Internals SA-240, Type 304 and Inconel-X Steam Generator Pressure Plate SA-212, Grade B Pressure Forgings SA-105, Grade II Cladding for Heads, Stainless SA-371, ER 308 Weld Rod SA-298, E 308 O-Cladding fer Tube Sheets Ni-Cr-Fe Tubes SB-163 (Code Case 1336)

Pressurizer Shell, Heads, and External Plate SA-212, Grade B Forgings and Nozzles SA-105, Grade II Cladding, Stainless Weld Rod SA-371, ER-308 SA-298, E 308 Internal Plate SA-240, Type 304 Internal Piping SA-312, Type 304 Piping 28 in. and 36 in.

Base Material to ASIM A-212, Grade B, or A-106, Grade C Clad on Inner Surface to ASIM A-371 Using Electrode Type ER-308 10 in.

AS'IM A-376, Type 316, and A-403, Grade WP 316 O

TABIE h-10 00 00337 4-34

]

Table 4-11 References For F16ure 4-4 Increase in Transition Temperature Due to Irraidation Effects For A302B Steel j

Neutron Ref.

Temp.,

Ixposure, IDIT, No.

Reference Material Type F

n/cm2 ( > l nev)

F 1

ASME Paper All Steels Max. Curve for 550.

Data No. 63-WA-100 (Figure 1).

2 ASTM-STP 380, A3TB Plate Trend Curve for 550.

Data P 295 18 3

NRL Report 6160, A302B Plate 550 5 x 10 65 p 12.

4 ASTM-STP 341, A302B Plate 550 8 x 1018 85(a) p 226.

5 ASTM-STP 341, A302B Plate 550 8 x 1018 100 p 226.

6 ASTM-STP 341, A302B Plate

$50 1 5 x 1019 130(*)

p 226.

7 ASTM-STP 341, A3TB Plate 550 1 5 x 1019 140 p 226, 8

Quarterly Report A302B Plate 550 3 x 1019 120 cf Progress,

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

1-31-65 19 9

Quarterly Report A302B Plate 550 3 x 10 135 of Frogress,

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

1-31-65 (a) Transverse specimens.

Oo TABIE h-11 00 00.438 3 55 L

Table 4-11 (Cont'd)

Neutron Ref.

Temp.,

Exposure, NDTT, No.

Reference Material Type F

n/cm2 ( > 1 mev)

F 10 Quarterly Report A302B Plate 550 3 x 1019 140 of Progress,

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

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

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

1-31-65 19 12 Quarterly Report A302B P3ste 550 3 x 10 205 of Progress,

" Irradiation Ef-fects on Reactor 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.

18 14, Welding Research A30GB Weld 500 5 x 10 So Supplement, Vol.

to 27, No. 10, Oct.

575 1962, p 465-S.

18 15 Welding Research A30es Weld 500 5 x 10 37 Supplement, Vol.

to 27, No. 1, Oct.

575 1962, p 465-S.

10 16 Welding Research A3CEB Weld 500

$ x 10 25 Supplement, Vol.

to 27, No. 10, Oct.

375 1962, p 465-S.

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y 7 Steam Temp./

Mean Shell 4

540 s

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Temp.

Shell Temp.

520 N

0 20 40 60 80 100 Bottom Top Tubesheet TUBE LENGTH, % OF TOTAL Tubesheet STEAM GENERATOR TEMPERATURES Q

EPont 000 NEE NUCLEAR STATION FIGURE 4-10

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WATER CONNECTIONS INLET AND OUTLET, TO COOLING JACKET FOR MOTOR BEARING Note: WATER JACKZ r ON MOTOR STATOR NOT SHOWN.

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T DISCHARGE SUCTION NOZZLE NOZZLE TYPICAL REACTOR COOLANT PUMP sura OCONEE NUCLEAR STATION FIGURE 4-11 c

000 351

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240 M5 W

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Years of Operation 80 l

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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 INTEGRATED NEUTRON EXPOSURE (E > 1 mev). n/cm x 10~ I9 PREDICTED NDTT SHIFT VERSUS REACTOR VESSEL IRRADIATION I

i OCONEE NUCLEAR STATION FIGURE 4-12 000 352 1