ML19319D681

Chapter 4 to Crystal River 3 & 4 PSAR, Rcs. Includes Revisions 1-10

ML19319D681
Person / Time
Site:	Crystal River, 05000303
Issue date:	08/10/1967
From:	FLORIDA POWER CORP.
To:
References
NUDOCS 8003240655
Download: ML19319D681 (56)

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TABLE OF CONTENTS Section Page 4 REACTOR COOLANT SYST_EM_ 41 4.1 DESIGN BASES 4-1 4.1.1 PERFORMANCE OBJECTIV E 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 Loads 4-2 4.1.2.4 Seismic Loads 4-2 4.1.2 5 Cyclic Ioads 4-2 4.1.2.6 Water Chemistry 4-2 4.1 3 EXPECTED OPERATING CONDITIONS k-2 4.1.4 SERVICE LIFE 4-3 4.1.4.1 Material Radiation Damage 4-3 4.1.4.2 Nuclear Unit Operational 'Ihermal Cycles 4-3 4.1.4.3 Operating Procedures 4-4 4.1.4.4 Quality Manufacture 4-5 4.1 5 CODES AND CLASSIFICATIONS 4-6 4.2 SYSTEM DESCRIPTION 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 Pressurizer 4-7 4.2.2 3 Steam Generator 4-8

( 4.2.2.4 Reactor Coolant Pumps 4-11 4.2.2 5 Reactor Coolant Piping 00000334 4-u 4-1

COFMNTS (Cont'd)

Section Page 4.2 3 PRESSURE-RELIEVING DEVICES 4-12 4.2.4 ENVIRONMENTAL PROHCTION 4-12 4.2 5 MATERIAIS OF CONSTRUCTION 4-12 4.2.6 MAX MUM HEATING AND COOLING RATES 4-14 4.2 7 LEAK DETECTION 4-14 43 SYSTEM DESIGN EVALUATION 4-16 431 SAFETY FAC'1VRS 4-16 4 3 1.1 Pressure Vessel Safety 4-16 4 3 1.2 Piping 4-21 4313 Steam Generator 4-21 432 RELIANCE ON INTERCONNECHD SYSTEMS 4-23 4.3 3 SYSTEM INTEGRITY 4-23 4.3.4 PRESSURE RELIEF 4-23 435 REDUNDANCY 4-24 436 SAFETY ANALYSIS 4-24 437 OPERATIONAL LIMITS 4-24 4.4 TEST AND INSPECTIONS 4-25 4.4.1 COMPONENT IN-SERVICE DISPECTION 4-25

.4.4.1.1 Reactor Vessel 4-25 4.4.1.2 Pressurizer 4-26 4.4.1 3 Steam Generator 4-26 4.4.1.4 Reactor Coolant Pumps 4-26 4.4.1 5 Piping 4-26 4.4.1.6 Dissimilar Metal and Representative Welds 4-26 h . - , -

4.4.1 7 Inspection Schedule 0000033b u-27 4-11

l CONTENTS (Cont'd)

Section Page 4.4.2 REACTOR C00IANT SYSTEM TESTS AND INSPECTIONS 4-27 4.4.2.1 Reactor Coolant System Precritical and Hot Leak Test 4-27 4.4.2.2 Pressurizing System Precritical Operational l Test 4-27 4.4.2 3 Pressurizer Surge Piping Temperature Gradient Test 4-27 4.4.2.4 Relief System Test 4-27 4.4.2 5 -Nuclear Unit Pbwer Startup Test 4-28 4.4.2.6 Nuclear Unit Power Heat Balance 4-28 4.4.2 7 Nuclear Nnit Power Shutdown Test 4-28 4.4 3 NATERIAL IRRADIATION SURVEILLANCE 4-28 Q 45 REFERENCES 4-30 b

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( LIST OF TABLES (At rear of Section)

Table No. Title Page 4-1 Tabulation of Reactor Coolant System Pressure 4-31 Settings 4-2 Reactor Vessel Design Data 4-31 4-3 Pressurizer Design Data 4-32 4-4 Steam Generator Design Data 4-32 4-5 Reactor Coolant Pump Design Data 4-33 4-6 Reactor Coolant Piping Design Data 4-34 4-7 Transient Cycles 4-34 4-8 . Design Transient Cycles 4-35

'4-9 Reactor Coolant System Codes and Classifications 4-35 4-10 Materials of Construction 4-36 4-11 Reference for Figure 4 Increase in Transition Temperature Due to Irradiation Effects for A302B Steel 4-37 m~

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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 4-6 Steam Generator Heating Regions 4-9 Steam Generator Heating Surface and Downcomer Ievel Versus Power

( 4-10 Steam Generator Temperatures

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4-11 Reactor Coolant Pump 4-12 Predicted NDTT Shift Versus Reactor Vessel Irradiation l

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4 REACTOR COOLANT SYSTEM 4.1 DESIGN BASES The_ reactor coolant system consists of the reactor vessel, coolant pumps, steam generators, pressurizer,.and interconnecting piping. The functional relationship between coolant system components is shown in Figure h-l.

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

! Piping and Valves - USASI B31.1-1955 (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 7essels.

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

d Reactor Vessel - ASME Boiler and Pressure Vessel Code, Gection III, Nuclear Vessels.

Welding Qualifications - ASME Boiler and Pressure Vessel Code, Section_IX.

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

4.1.1 PERFORMANCE OBJECTIVES The reactor coolant system is designed to contain and circulate reactor coolant at pressures and flows necessary to transfer the heat generated in the reactor core to the secondary fluid in the steam generators. In addition to serving as a heat transport medium, the coolant also serves as a neutron moderator and reflector, and as a solvent for the soluble poison (boric acid) utilized in chemical shim reactivity control.

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

4.1.2 DESIGN CHARACTERISTICS 4.1.2.1 Design Pressure 00000339 y

(E 'The reactor coolant system design, operating, and control set point pres-sures'are' listed in Table 4-1. The design pressure allows for operating 4-1

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transient pressure changes. The selected design margin considers core .

themal lag, coolant transport times and pressure drops, instrumentation and control response characteristics, and system relief valve character-istics. The design pressures and data for the respective system compo-nents are listed in Tables 4-2 through 4-6.

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

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

4.1.2.4 Seismic Loads Reactor coolant system co=ponents are designated as Class I equipment, and are designed to maintain /$E X functional integrity during earthquake.

The basic design guide for tee seis=ic analysis is the AEC publication TID-7024, " Nuclear Reactors and Earthquake". Structures and equip =ent vill be designed in accordance with Appendix 5A.

h 4.1.2 5 Cyclic Ioads All components in the reactor coolant system are designed to withstand the effects of cyclic loads due to reactor 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 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 reactivity control and to minimize corrosion of reactor coolant system surfaces. The reactor coolant chemistry is discussed in further detail in 9 2.

4.1 3 sawrw OPERATING CONDITI0IiS Throughout the load range from 15 to 100 per cent power, the reactor cool- .

ant system is operated at a constant average temperature. Reactor coolant system pressure is controlled to provide sufficient overpressure to main-tain adequate core subcooling.

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()f' The minimum operating pressure is established from core thermal analysis.

This analysis is based upon the maximum expected inlet and outlet temper-atures, the maximum reactor power, the minimum DNER required (including instrumentation errors and the reactor control system deadband), and a core flow distribution factor. The maximum operating pressure is estab-lished 'en the basis of ASME Code relief valve characteristics and the margin: required for nomal pressure variations in the system. Pressure control between the preset maximum and minimum limits is obtained di-rectly by pressurizer spray action to suppress high pressure and pres-surizer heater action to compensate for low pressure. Nomal operational lifetime transient cycles are discussed in detail in 4.1.4.

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 established operating procedures. In the following dis-cussion each of these life-dependent factors will be discussed with re-gard to the affected components.

4.1.4.1 Material Radiation Damage l The reactor vessel is the only reactor coolant system component exposed to a significant level of neutron irradiation and is therefore the only l p component subject to material radiation damage. To assess the potential U radiat'.cn damage at the end-of-reactor service life, the maximum exposure l

l from feJt neutrons (E > 1.0 Mev) has been computed to be 3 0 x 10 19 n/cm2 ovcc a 40 year life with an 80 per cent load factor. Reactor vessel ir-radiation exposure calculations are described in 3 2.2.1 7 For this neutron exposure, the predicted Nil-Ductility Transition T er-ature (NUIT) shift is 250 F based on the curve shown in Figure 4-4. 1 Based on en initial NUIT of 10 F, this shift vould result in a predicted j NDIT .of 260 F.

The " Trend Curve for 550 F Data", as shown in Figure 4-4, represents ir-radiated ca*,erial test results and was compiled from the reference docu-ments listed in Table h-11.

To evaluate the NDTT shift of velds, heat-affected zones, and base mate-rial for the ma'erial used in the vessel, test coupons of these three material types have been included in the reactor vessel surveillance l program as described in h.k.3 4.1.4.2 .

Nuclear Unit Operational Thermal Cycles 00000341 1

To establish the service life of the reactor coolant syster components as required by the ASME III for Class "A" vessels, the nuclear unit operating conditions that involve the cyclic application of loads and thermal con-ditions have been established for the h0-year design life.

The number of themal and loading cycles to be used for design purposes are listed in Table h-7, " Transient Cycles". The estimated actual cycles

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based on a review of existing nuclear stations operations are also provided in Table 4-7 Table 4-9 lists those components designed to ASME III - Class "A".

The effect of individual transients and the sum of these transients are evalu-h ated to determine the fatigue usage factor during the detail design and stress analysis effort. As specified in ASME III Paragraph 415 2 (d)(6), the cumula-tive fatigue usage factor vill be less than 1.0 for the design cycles listed in Table 4-7 The transient cycles listed in Table h-7 are conservative and complete in that they include all significant modes of normal and emergency operation. The es-timated frequency bases for the design transient cycles are listed in Table 4-8.

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 chan6e 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 ncr-mal refueling operation only the reactor vessel closure seals are hydrotested for pressure integrity; therefore, reactor coolant system hydrotesta before startup are not included.

4.1.4.3 operating Procedures The reactor coolant system pressure vessel components are designed using the transition temperature method cf minimizing the possibility of brittle fracture of the vessel materials. The various co=binations 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 perfor=ed in agegrdance with a " Frac-ture Analysis Diagram" as published by Pellini and Puzak.(2j At temperatures below the Nil-Dactility Transition Temperature (NDTT) and the Design Transition Temperature (DTT), which is equal to NDIT + 60 F, the pressure vessels will be operated so that the stress levels vill 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 maximu= 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 no=.'nal stesses are held within the referenced stress limits (a through c above), brittle f a Thi tatement is based on data re-portedbyRobertson3$turevillnotoccur.

and Kihara and Masubichi in published literature. It can be shown that stress li=1ts can be controlled by i= posing operating procedures 00000342

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s that control pressure and temperature duriq 'leatup and cooldown.(3) This procedure vill insure that the nominal stress levels do not exceed those specified in a through e above.

4.1.4.4 Quality Manufacture i

Material selection is discussed in detail 1: 4.2 5 After receipt of the material a program of qualification of all' welding and heat treating processes that could affect mechanical or metallurgical properties of the material during fabrication is undertaken. This pro-gram vill establish the properties;of the material, as received, and certify that the mechanical properties of the materials in the finished vessels are consistent with those used in the design u alysis. This pro-gram consists of:

1

a. Weld qual'.7; cation test plates using prodtw hen procedures and t

. subjecting test plates to the heat treatments to be used in fabricating the vessels.

b. . Subjecting qualf.fication test plates to all nondestructive tests to be employed in production, such as x-ray, dye-penetrant, ma6netic particle, and ultrasonic. Acceptance standards are the same as used for production.

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

(2) Ductility.

(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 inspection procedures are prepared.

t 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 of these dimensions is kept for future reference.

I All velders are qualified or requalified as necessary in t:.cordance with The' Babcock & Wilcox Company and ASME IX requirements. Mah lot of weld-

.ing electrodes and fluxes is tested and qualified before release to in-sure that required mechanical properties and as-deposited chemical prop-l erties can be met. Electrodes-are identified and issued only on an ap-L proved request to_ insure that the correct materials are used in each veld.

All velding electrodes and fluxes are maintained dry and free from contami-nation before use.: Records are maintained and reviewed by velding engineers

' (A ' to insure that approved procedures and materials are being used. Records

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are maintair.ed for each veld joint and include the velder's name, essen-tial veld perameters, and electrode heat or lot number.

The several types of nondestructive tests perfomed during vessel fabrica-tion are as follows:

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 specifications require,

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

c. Magnetic Particle Examination is used to detect surface or near surface defects in machined veld grooves prior to velding, com-pleted veld surfaces, tmd the comple w external surface of the vessels including veld seams after fint heat treat =ent.

d. Liquid Penetrant is used to detect surface def'ets in the veld deposit claddin6, nonmagnetic materials, and closure studs.

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

4.1 5 CODES AND CLASSIFICATIONS All pressure-containing components of the reactor coolant system are de-signed, fabricated, inspected, and tested to applicable codes as listed in Table 4-9 4.2 SYSTEM DESCRIPTION AND OPERATION j 4.2.1 GENERAL DESCRIPTION l

l The reactor coolant system consists of the reactor vessel, two vertical once-through steam generators, four shift-sealed coolant circulating pumps,- an electrically heated pressurizer, and interconnecting piping.

The system is arranged as two heat transport loops, each with two circu-lating pumps and one steam generator. Reactor system design data are listed in Tables 4-2 through 4-6, and a system schematic diagram is shown in Figure 4-1. Elevation and plan views of the arrangement of '.Ne major components are shown in Figures 4-2 and 4-3 4.2.2 MAJOR COMPONDTIS 4.2.2.1 Reactor Vessel 00000344 The reactor vessel consists of a cylindrical shell, a cylindrical support skirt, a spherically dished bottom heod, and a ring flange to wh$ch a re-movable reactor clonure head is bolted. The reactor closure head is a g spherically dished head velded to a ring flange. W

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q h The vessel has six major nozzles for reactor coolant flow, 69 control rod drive nozzles and a vent line nozzle mounted on the reactor closure head, and two core flooding nozzles--all locateti above the core. The vessel closure seal is formed by two concentric O-rings with provisions between them for leak detection. The reactor vessel, nozzle design, and seals in-corporate the extensive design and fabrication experience accumulated by MW. Fifty-two incore instru=entation nozzles are located on the lover head.

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

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

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

The reactor vessel shell material is protected against fast neutron flux and ga=ma heating effects by a series of water annull and the thermal b

W shield located between the core and vessel vall. This protection is fur-ther described in 3 2.hol.2, 4.1.4, and 4 31.

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

Surveillance specimens made from reactor steel are located between the reactor vessel vall and the thermal shield. These specimens vill be ex-smined at selected intervals to evaluate reactor vessel material NDIT changes as described in h.h.3 -

The reactor vessel general arrangement is shown in Figure 4-3, and the general arrangement of the reactor vessel and internals is shown in Fig-ures 3 43 and 3-46. Reactor vessel design data are listed in Tabic h-2.

4.2.2.2 f Pressurizer The generall 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 l

and a water reserve to acco=modate reactor coolant volume changes during operation. t The pressurizer is a vertical cylindrical vessel connected to the reactor outlet piping by the surge piping. The pressurizer vessel is protected 4-7 -

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from themal effects by a thermal sleeve on the surge line and by a dis-tribution baffle located above the surge pipe entrance to the vessel. h 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 discuased in 4 3 4.

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

The pressurizer contains replaceable electric heaters in its lower sec-tion and a water spray nozzle in its upper section to maintain the steam and vater at the saturation temperature corresponding to the desired re-actor coolant system pressure. During outsurges, as the pressure in the reactor decreases, some of the water in the pressurizer flashes to steam to maintain pressure. Electric heaters are actuated to restore the nor-cal operating pressure. During insurges, as pressure in the reactor sys-t?m increases, steam is condensed by a water spray from the reactor inlet lines, thus reducing pressure. Spray flov and heaters are controlled by the pressurizer pressure controller.

Instrumentation for the pressurizer is discussed in 7 3 2.

4.2.2 3 Steam cenerator The general arrangement of the steam generators is shown in Figure 4-7, i and design 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 through the tubes, and steam is generated en the shell side. The high pressure parts of the un}t are the hemisphe-rical heads, the tubesheets, 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 l

of the steam-producing section of the vessel. Within the shell, the tube l bundle is surrounded by a baffle, which is divided into two sections.

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

I Vents, drains, anstrumentation nozzles, and inspection openings are pro-vided on the shell side of the unit. The reactor coolant side has in-l strumentation connections on the top and bottom heads, manvays on both l heads, and a drain nozzle for the bottom head. Venting of the reactor Inconel is a trade name of an alloy manufactured by the International Nickel Company. It also has substantial common usage as a generic de-scription of a Ni-Fe-Cr alloy conforming to ASTM Specificati n ,S - 63 It is'in the latter context that reference is made he QQ p g-

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/ coolant side of the unit is accomp .ished by a vent connectAon on the re-actor ceolant inlet pipe to each unit. The unit is suppcrted by a skirt attached to the bottom head.

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

Four heat transfer regions exist in the steam generator as feedvater is ,

converted to superheated steam. Starting with the feedvater inlet these are

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

The saturated water in the downcomer for=s a static head to balance the static head in the nucleate boiling section. This n provides the head to overcome pressure drop in the circuit V formed by the downcomer, the boiling sections, and the bypass l steam flow to the feedvater heating region. With low (less than1ft/sec)saturatedwatervelocitiesenteringthegener-ating section, the secondary ~ side pressure drops in the boil-ing section are negligible. The majority of the pressure drop is due to the static head of' the mixture. Consequently, the downcomer level of water balances the mean density.of the two -

phase boiling mixture in the nucleate boiling region.

i b. Nucleate Boiling The s,aturated water enters the tube bundle, and the steam-water mixture flows upward on the outside of the Inconel tubes counter-current to the reactor' coolant flow. The vapor content of the mixture increaser almost uniformly until DNB, i.e. , departure of nucleate boiling, is reached, and then film boiling and super-heating occurs. The quality at which transition from nucleate boiling w 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 l

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' 0000047 q, Saturated steam is raised to final temperature Tn the super-

" *-heater region.

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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 flov from 15 - 100 per cent load. A constant minimum level is held belov 15 per cent load. The amount of surface (or length) of the nucleate boiling section and the film boiling section is proportional to load. The sur-face available for superheating varies inversely with load, i.e., as load decreases the superheat section gains from the nucleate and film boiling regions.

Mass inventory in tne 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 botn tne 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 effect of the mixture. The boiling in the steam generator is somewhat similar to " pool boiling" except that there is motion upward that permits some parallel flow of water and steam.

A plot of reactor coolant and steam temperatures versus power is shown in Figure 7-5 As shown, both steam pressure and average reactor coolant temperature are held constant over the load range from 15 to 100 per cent full power. Constant steam pressure is obtained by a variable two-phase boiling length (see Figure 4-8) and by the regulation of feed flow to ob-tain proper steam generator secondary mass inventory. In addition to average reactor coolant temperature, reactor coolant flow is also held g constant. The difference between reactor coolant inlet and outlet tem- W peratures increases proportionately as load is increased. Saturation pressure and temperature are constant, resulting in a variable superheater outlet temperature.

Figure 4-10, a plot of te=perature versus tube length, shows the tempera-ture differences between shell and tube throughout the steam generator at full load. The excellent heat transfer coefficients pemit the use of a l

secondary operating pressure and te=perature sufficiently close to the reactor coolant average temperature 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 sat-urated water from feedvater inlet to the lover tubesheet.

In the superheater section, the tube vall te=perature approacl.as the re-actor coolant fluid te=perature since the steam film heat transfer coef-ficient is considerably lover than the reactor coolant heat transfer coef-ficient. By baffle arrangement in the superheater section, the shell sec-tion is bathed with superheated steam above the steam outlet nozzle, fur-ther reducing temperature differentials between tubes and shell.

The steam generator design and stress analysis will be performed in ac-cordance with the requirements of the ASME III as described in 4 3 1.1. .

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h.2.2.4 Reactor Coolant Pumps 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 suc-tion and horizontal discharge. Each " ump has a separate, single-speed, top-mounted motor, which is connecte- to the pu=p by a shaft coupling.

Shaft sealing is accomplished in tr part of the pump housing using a throttle bushing, a seal chamber, - - 4 si seal, and a drain chamber in series. Seal water is injected ahs m a throttle bushing at a pressure approximately 50 psi above reactor system prescure. Part of the seal flow passes into the pump volute through the radial pump bearing.

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

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

(~} housing. An oil-lubricated radial bearing and a Kingsbury type, double-b/ acting, oil-lubricated 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 lub-rication is maintained by internal pumping provisions in the motor, or by and external system if required for " hydraulic-jacking" of the bearing surfaces for startup.

An antirotation device will be furnishec with each pump motor to prohibit l reverse rotation of the pump.

I Factory thrust, vibration, and seal performance tests will be made in a closed loop on the first pump at rated speed with the pump end at rated temperature and pressure. Sufficient testing will be done on subsequent units to substantiate that they conform to the initial test pump charac-teristics.

4.2.2 5 Reactor Coolant Piping l 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 piping connection, the piping is t

equipped with welded connection; for ;ressure taps, temperature elements, i vents, drains, decay heat removal, and emergency core cooling high pres-

! /sure injection water. Thermal sleeves are provided in the pressurizer j syrge piping and the emergency high pressure injection pipe connections.

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4.2 3 PRESSURE-RELIEVING DEVICES g

The reactor coolant system is protected against overpressure by control and protective circuits such as the high pressure trip and code relief valves located on the top head of the pressuriser. The r311ef valves discharge into the reactor coolant drain tank which condenses and collects the effluent. The schematic arrangement of the relief devices is shown in Figure 4-1. Since all sources of heat in the system, i.e., core, pres-surizer heaters, and reactor coolant pumps, are interconnected by the re-actor coolant piping with no intervening isolation valves, all relief pro-tection can conveniently be located on the pressurizer.

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

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

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

4.2 5 MATERIALS OF CONSTRUCTION Each of the materia 2s 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-resistut materials consisting of 304 or 316 ss, veld deposit 304 ss clad-ding, Inconel (Ni-Cr-Fe), and 17-4 PH (H1100). These materials were chosen i

for specific purposes at various locations within the system because of their superior compatibility with the reactor coolant.

Periodic analyses of the coolant chemical composition vill be performed 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 corro-sion 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 preven. deposits and corrosion inside the

steam generators. This required feed nter quality has been successfully l used in comparable once-through, nonnuclear steam generators. The phe-nomena of stress-corrosion cracking and corrosion fatigue are not gener-ally encountered unless a combination of elements in varying degrees is g

4-12 00000350

present. The necessary conditions are a susceptible alloy, an aggres a e

/ environment, a stress, and time.

All external insulation of reactor coolant system components vill be com-patible with the component mt.Lerials. 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 st.rfaces in the reactor coolant system vill be insulated with lov or halide-free insulating mate-rial as required.

The reactor vessel plate material opposite the core is purchased to a specified Charpy V-notch test result of 30 ft-lb or greater at a corre-sponding nil-ductility transition temperature (NDIT) of 10 F or less, and the material vill be tested to verify conformity to specified re-quirements and to determine the actual NDIT value. In addition, this plate vill be 100 per cent volumetrically inspected by ultrasonic test using both nomal and sh' ear wave.

The reactor vessel material is heat-treated specifically to obtain good notch-ductility which vill insure a low NDIT and thereby give assurance '

that the finished vessel can be initially hydrostati: ally tested and op-erated at 2..om temperature without restrictions. The stresa limits es-j tablished for the reactor vessel are dependent upon the temperature at p which the stresses are applied. As a result of fast neutron absorption V in the region of the core, the material ductility will change. The effect is an increase in the NDIT. The predicted end-of-life NDIT value of the reactor vessel opposite the core is 260 F cr less. The predicted neutron exposure and NDIT shift are discussed in 4.1.4.~

l The unirradiated or initial NDIT of pressure vessel base plate material i' presently measured by two methods: the drop veight test given in ASTM E208, and the Charpy V-notch impact test (Type A) given in ASTM E23 The NDIT is defined in ASTM E208 as "the temperature at which a specimen is

( broken in a series of tests in which duplicate no break performance occurs at a 10 d higher temperature". Using the Charpy V-notch test, the NDIT is defined as the temperature at which the energy required to break the

! specimen is a certain " fixed" value. For SA 302B steel the ASME III l Table N-332 specifies an energy value of 30 ft-lb. This value is based on a correlation with the drop veight test and will te referred to as the "30 ft-lb fix". A curve of the temperature versus energy absorbed in breaking the specimen is plc,tted. 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 NDIT.

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

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The material for these testa vill be treated by the methods outlined in ASME III Paragraph N-313 The test coupons will be taken at a distance oft /4(1/4oftheplatethickness)fromthequenchedsurfacesandata h distance of T from the quenci:ed edges. These tests are perfomed by the material supplier to certify the material as delivered to B&W. The exac+

test coupon locations are reviewed and approved by B&W to insure compli ance with the applicable ASME Code and specifications. In accordance with ASME III Paragraphs N-712 and N-713, B&W perfoms Charpy V-notch impact tests on heat-affected zone (HAZ), base metal, aad weld metal on all pressur- ressel test plates.

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

The reactor vessel design includes surveillance specimens which will per-mit 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 cool-ant system components are purchased to the appropriate design code re-quire =ents and specific component function.

The material irradiation surveillance program is described in 4.4 3 4.2.6 MAXIMUM 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 cool-ingrateis100F/hr. The exact final rates are detemined during the detail design and stress analysis of the vessel.

The fastest cooldovn rates resulting from the break of a main steam line are discussed in l?+.1.2 9 4.2 7 LEAK DETECTION To minimize leakage from the reactor coolant system all components are interconnected by an all-velded piping system. Some of the components have access openings of a flanged-gasketed design. The largest of these is W. reactor closure head, which has a double metal 0-ring seal with provirions for monitoring for leakage between the 0-rings. Other open-ings. uld appurtenances to the reactor coolant system that are possible sources of leakage are tabulated in detail along with the maximum ex-pected rates of leakage in Section 11.

With regard to the reactor vessel, the probability of a leak occurring is considered to be remote on the basis of reactor vessel design, fabrication, test, inspection, and operation at temperatures above the material NInT 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 transier' operation, such as an emergency cooldown, would there be 4 -

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some leakage past the innermost 0-ring seal. A stress analysis en a sim-

.d ilarvesseldesignindicatesthisleakratewouldbeapproximately10cc/

J min through the sest monitoring taps to a drain,'and no leaka6e will occur past the outei- 0-r .ng seal. The exact nature of this transient condition and the resulting small leak rate will be detemined 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 will be i

detected by one or more of the following methods:

a. Instrumentation in the control room will indicate the addition rate of makeup water required to maintain normal water level in the pressurizer and in the makeup tank. Deviation from nomal makeup and letdown to the reactor coolant system will provide l1 an indication of the ma6nitude of the leak.

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

{ c. Control room instrumentation will indicate the existence of a change in the water level in the reactor building sump.

If any one of the methods above indicates an excessive reacter coolant leakage rate during operation, the reactor will be taken to a6 cold shut-6.own, and the cause of the problem will be determined.

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43 SYSTEM DESIGN EVALUATION 431 SAF Tl FACTORS The reactor coolant system is designed, fabricated, and erected in accor-dance with proven and recognized desiG n codes and quality standards appli-cable for the specific component function or classification. These com-ponents are designed for a pressure of 2,500 psig at a nominal tempera-ture of 650 F. The corresponding nominal operating pressure of 2,185 psig allows an adequate margin for normal load changes and operating transients. The reactor system components are designed to meet the codes listed in Table 4-9 Aside from the safety factors introduced by code requirements and quality control programs, as described in the following paragraphs, the reactor coolant systen functional safety factors are discussed in Sections 3 and 14.

4 3 1.1 Pressure Vessel Safety The safety of the nuclear reactor vessel and all other reactor coolant system pressure vessels is dependent upon four major factors: (1) de-sign and stress analysis, (2) quality control, (3) proper operation, and (4) relief valves. The special care and detail used in implementing these factors in pressure vessel manufacture are briefly deceribed as follows:

4.3 1.1.1 Design and Stress Analysis g

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 from industry and government who are skilled in the design and fabrication of pressure vessels. It is a comprehensive code based on the most appli-cable stress theory. It requires a stress analysis of the entire vessel under both steady-state and transient operations. The result is a com-plete evaluation of both primary and secondary stresses, and the fatigue life of the entire vessel. This is a contract with previvus codes which basically established a vescel thickness during steady-state operations only. In establishing the fatigue life of the e pressure vessels, using the design cycles from Table 4-7, the fatigue evaluation etcves 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 l played a major part in refining these analytical solutbns. The Babcock

& Wilcox Company has confirmed the theory of plates and shells by measur-ing strains and rotations on the Isrge flanges of actual pressure vessels and , finding them to be in agreement with those predicted by the theory.

B&W 'haa' alsh coriducted laboratory deflection studies of thick shell and ring combinations to define the accuracy of the theory, and is using com-puter programs developed on the basis of this test data.

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1 0 The analytical procedure considers all process operstion conditions. A detailed design and analysis of every part of the vessel is prepared as follows:-

a. The vessel size and configuration are set to meet the process requirements, 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., re-quired.by ASME III.

I 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. Computer programs are used in this development.

c. The stresses thrcugh 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 Inmbrane stress (to insure adequate tensile strength of the vessel), membrane plus primary bending stre m (to insure a. dis-tortion-free vessel), secondary stress (to insure a vessel that will not progressively deform under cyclic loading), and peak stresses (to insure a vessel of maximum fatigue life).

l A design report is prepared and submitted to the Jurisdictional.authori- ,

( ties and regulatory agencies, i.e., state, insurance, etc. This report defines in sufficient detail the design basis, loading conditions, etc.,

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

4 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 maintained to keep all measuring equipment in proper calibration, and personnel supervising this work are trained in formal programs spon-

, sored by gage equipment manufacturers.

l The practice of applied radiography is being continually improved to en-hance flaw detection. Present procedures are:

a. All welds c.re properly prepared by chipping and grinding valleys l

between stringe beads so that radiographs can be properly in-terpreted. .

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

c. An 0.080 in lead filter is used at the film to absorb " broad-beam scatter" when using high voltage equi 1xnent (above 1 Mev).

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

e. Densities of radiographs are controlled by densitometers,

f. Double film technique is used on all gam =a-ray exposures as well as high voltage exposures.

g. Films are processed through an automatic processor which has a controlled replenishment, temperature, and process cycle, all contributing to better quality.

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

Ultrasonics, one of the most useful of inspection tools, is being used as follows:

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

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

i c. The bond between cladding and base material is inspected by l ultrasonics.

i

! d. All plate is 100 per cent volumetrically inspected by ultrasonics using 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 employed on both parts and finished vessels as follows:

a. Welds are inspected with the =agnetic particle method after re-moval of backup strips.

b. Weld preparations are inspected by the ragnetic particle m;thod.

c. Tne external surface of the er.. ire vessel, including veld seams, is inspected after all heat treatment.

d. Personnel using this method are trained by d&W and by the equip-ment manufacturers who offer fomal training programs.

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?> The liquid penetrant examination helps to detect surface defects and is particularly adaptable to the non=agnetic materials 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 using this method of exa=ination are trained by B&W and by the equipment manufacturers.

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

The size of defect that can conceivably contribute to the rupture of a vessel depends not only on the size effect but on the orientation of the defect, the magnitude of the stress field, and te=perature. These major parameters have been correlated by Pellini and Puzak(2) who have prepared a " Fracture Analysis Diagram" which is the basis of vessel operation from cold startup and shutdown to full pressure and te=perature operation.

The diagram predicts that, for a given level of stress, larger flav sizes vill be required for fracture initiation above the NDTT temperature. For example,atstressesintheorderof3/4yieldstrength,aflawinthe order of 8 to 10 in. may be sufficient to initiate fracture at temperatures below the NDTT temperature. However, at NDTT -F 30 F, a flav of 1-1/2 ti=es this size may be required for initiation of fracture. While at a tempera-ture of I&l'T + 60 F, brittle fracture is not possible under elastic stresses because brittle fracture propagation does net take place at this temperature. Fractures above this temperature are of the predominantly ductile type and are dependent upon the member net section area and sec-tion 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 vill be loss-of-back-reflection greater than that from a 1/2 in.

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

The stress analysis of the studs will include a fatigue evaluation. It is ,

not expected that fatigue evaluation vill yield a significantly high usage factor for the 40-year design life. Therefore, there vill be no planned frequency for stud replace =ent. If en indication is found when the studs are inspected during refueling, as desevibed below, the stud vill be re-placed.

One-third of the studs vill be visually examined and dye-penetrant in-q spected at every refueling. Any positive indications found will be cause (j for rejection. ,

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The reactor closure head is attached to the reactor vessel with sixty 6- ~~

l 1/2in.diameterstuds. The stud material is A-540, Grade B23 (ASME III, -@ 1 Case 1335) which has a minimum yield strength of 130,000 psi. 'Ihe studs , l when tightened for cperating conditions, vill haw a tensile stress of approximately 30,000 psi. Thus, at operating conditions (2,185 psig):

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

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

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

d. 43 sp= metrically located studs can fail before the remaining l studs reach yield strength.

4 3 1.1 3 operation As previously mentioned in 4.1.4, pressure vessel service life is depen-dent on adherence to established operating procedures. Pressure vessel safety is also dependent on prcper vessel operation. Therefore, particu-lar attention is given to fatigue evaluation of the pressure vessels and to the factors that affect fatigue life. The fatigue criteria of ASME 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 ves- g sels are dependent upon the te=perature at which the stresses are applied. W As a result of fast neutron absorption in the region of the core, the re-

, actor vessel material ductility will change. The effect is an increase l in the nil-ductility transition temperature (UDTI). The determination of l the predicted NDTT shift is described in 4.1.4.1. This NDTT shift is factored into the p) ant startup and shutdown procedures so that full oper-ating pressure is not attained until the reactor vessel temperature is l

above the design transition temperature (DTT). Below the DTT the total

! stress in the vessel vall due to both pressure and the associated heatup and cooldown transient is restricted to 5,000 - 10,000 psi, which is be-lov the threshold of con:ern for safe operation. These stress levels de-fine an operating coolant pressure temperature path or envelope for a stated heatup or cooldown rate that must be followed. Additional infor-cation 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 previoucly mentioned in 4 31.1 above but which are considered conserva-tive and provide an additional margin of safety, are as follows:

a. Use of a stress concentration factor of 4 on assumed flaws in calculating stresses.

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

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

d. The design shift in ND'IT as given in 4.1.4.1 is based on maxi-mum predicted flux levels at the reactor vessel inside vall sur-fe.ce, whereas the bulk of the reactor vessel material will ex-perience a lesser exposure of radiation and consequently a lover change in NDTT over the life of the vessel.

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

f. Results from the method of neutron flux calculations, as de-scribed in 3 2.2.17, have increased the flux calculations by a factor of 2 in predictus the nv*t 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

j. De foregoing discussion presents a detailed description of quality design, fabrication, inspection, and operating procedures used to insure con'idence in the integrity of pressure vessels. Experience reported by Reference (5),

E&W, and the satisfactory experience of B&W customers support the conclu-sion that pressure vessel rupture is incredible.

4.3 1.2 Piping Total stresses resulting from thermal expansion and pressure and mechani-cal and seismic loadings are all considered in the decign of the reactor coolant piping. The total stresses that can be expected in the piping are within the =y4=m code allowables. Be pressuricer surge line con-nection and the high pressure injection connections are equipped with thermal sleeves to limit stresses from thermal shock to acceptable values.

All materials and fabrication procedures will meet the requirements of the specified code. All material vill be ultrasonically inspected. All velds will be radiograph 1cally inspected. All interior surfaces of the inter-connecting piping are clad with stainless steel to eliminate corrosion problems and to reduce coolant contamination.

4.3 1 3 Steen Generator

' Because the basic concept of the once-through steam generator would indi-

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themselves are designed to take the full design pressure on either side j of the tubesheet 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 this design i premise in a transient manner. Secondary temperature variations during the accident e.re essentially transient skin effects with the controlling temperature for the tubesheets and tubes being that of the reactor cool- I ant. Thermal stresses for this case vill be below ASME allowable values.

Some tube deformation may occur but vill be restrained by the tube sup-ports.

During nomal power operation the tubes are hotter than the shell of the steam generator by 10 to 20 F depending on load. 'Ihe effect is to put the tubes in a slight ccepression of 3,000 psi at the 20 F maximum tem-perature difference.

During startup and shutdown operations the tubes are botter than the shell of the steam generator by 40 F. This places the tubes in a compressive ^

stress of 6,000 psi. Thus, the stress levels develeped during norma 3 startup or shutdown operations cause no adverse effect on the tubes since these stresses are well below tr.e allowable stress of 23,300 psi for this SB-163 material (ASME III, Case 1336). Buckling of the tubes dces not occur since they are supported laterally at 40-in. intervals along their length. To de=onstrate the structural adequacy of the steam generator at this condition, a laboratory unit was constructed of the same tube size, g length, and material as the steam generator, but of seven tubes in nu=ber. W 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 performed with a tube-to-chell temperature difference twice as great as the maximum ex-pected during startup and shutdown (Transients 1 and 2, Table 4-7). De-structive exa=ination of.the unit after this test indicated no adverse effects frcxn fatigue, stress, buck. ling, or tube-to-tubesheet joint leak-age.

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1 Mav) 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 1010 65 p 12 4 ASTM-STP 341, A302B Plate 550 8 x 1010 85(a) p 226 5 ASTM-STP 341, A302B Plate 550 8 x 1010 109 p 226 6 ASTM-STP 341, A302B Plate 550 1 5 x 1019 130(*) p 226 7 ASTM-STP 341, A302B Plate 550 1 5 x 10 19 140 p 226 8 quarterly Report A302B Plate 550 3 x 1019 120 of Progress, " Irradiation Ef-fects on Reactor Structural Mate-rials", n-1-64/ 1-31-65 9 Quarterly Report A302B Plate 550 3 x 1019 135 of Progress, " Irradiation Ef-fects on Reactor Structural Mate-rials",11-1-64/ &wil-31-65 (") Transverse speciment. hh/} A ' Q) ,c TABLE 4-11 4-37 Table k-11 (Cont'd) g Neutron Ref. Temp., Exposure, NDTT, No. Reference Material Type F n/c=2 (> 1 Mev) F lo Quarterly Report A302B Plate 550 3 x 1o19 14o of Progress, " Irradiation Ef-fects on Reactor Strt.ctural Mate-rials",11-1-64/ l-31-65 11 Quarterly Report A302B Plate 550 3 x 1o19 17o of Progress, " Irradiation Ef-fects on Reactor Structural Mate- - rials",11-1-64/ 1-31-65 12 Quarterly Report A3o2B Plate 550 3 x 1019 205 of Progress, " Irradiation Ef-fects on Reactor Structural Mate-rials",11-1-64/ & W 1-31-65 13 Welding Research A302B Weld 500 5 x 1018 7o Supplement, Vol. to 27, No. 12, oct. 575 1962, p 465-S 14 Welding Research A302B Weld 500 5 x 1o18 So Supplement, Vol. to 27, No. lo, oct. 575 1962, p 465-S 15 Welding Research A302B Weld soo 5 x 1o18 37 Supplement, Vol. to 27, No. 1, Oct. 575 1962, p 465-S 16 Welding Research A302B vid 500 5 x 1o18 25 Supplement, Vol. to 27, No. lo, oct. 575 1962, p 465-S 7 QQkJ TABLE 4-11 (Cont'd) -t 4-38 9 i Gft sf r AC s f N T TC R C C M &sif "A % h Q:RD $AMritMp-?C C A - Fo I /1% 4~Af 55uRi/ER w arm 2195 951 WE N T= TO WO , a EZD CEC A r hfAT AC M 's AL-TOON 'I b 7E AM GE AE A ATOR S TE A M e-- 5 TEAM - FEEC wA TER - I 5EAL vEhr ~C AD ~ == Q MN DRJ)N-T:89 0 ~ ~ * ~ sc + - -- o-.1: $EAL n A TER ~~-~ e-$Er nATER AETURWTOMU sfMnCM ht V 1 d, CGA> N / h #' # # S A VCl o h 5 - l TCwa$CA , ORAnv-ro no l ' $fAL VEN Y-Tv WD w-fEAs CM A.N TO WD o---- IC w ~~~ e- AC EfAL WA 7[R** ** ~

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&Docket f2 J3 no2. 50-192

-bf

._ '~ 50-393 February 7,1%8

(%,, ,,

V AMENDMENT NO. 2 FLORIDA POWER CORPORATION Crystal River Plant Units 3 & h Amendment No. 2 to the Florida Power Corporation's Preliminary Safety Analysis Report includes both insert pages, as listed below, and Supplement No. 1.

Supplement No.1 censists of the response to the letter dated January 19, 1968 from Dr. Peter A. Morris, Director, Division of Reactor Licensing. The supple-ment also contains answers to the " Requested Additional Infomation" by D.R.L.

on January 19, 1968 (1.0 through 9.15).

NOTE Inclemed herewith, immediately following the instruction sheets, you will find a title sheet entitled " Appendices" and a complete Table of Contents (pages "i" through "vii"). These sheets are to be inserted in the binder entitled " APPENDICES which accompanies this amendment.

Remove the contents in existing Volume h--from the tab entitled " APPENDIX" to the end of the volume (Appendix 12A)--and insert all annendices in the binder entitled

" APPENDICES." (The title sheet and table of contents in existing Volume h should remain in place.)

The follew!ng sheets of the Florida Power Corporation's Preliminary Safety Analysis Report are to be deleted and, where appropriate, revised sheets dated 2-7 68 should be inserted.

Remove the fc11cving sheets: Insert the folleving sheets :

Paves: Table of centents - vii Pages : Table of contents - vii (one each in Volumes 1, 2, 3, and h) (one each in Volumes 1, 2, 3, and h)

Pages: 1-27, 1-28, 1-29, 1-30, 1-35, Pages: 1-27, 1-28, 1-29, 1-30, 1-35, ar.d 1-36. and 1;36.

Fiaures: 1-2, 1-3, 1-4, 1-5, 1-6, 1-7, Figures: 1-2, 1-3, 1 h, 1-5, 1-6, 1-7, 1-9, and 1-12. 1-9, and 1-12.

Pages: 5-17, 5-18, and 5-19. Pages: 5-17, 5-18, and 5-19. l 1

Figure: 5-6. Figure: 5-6.

1 ns \

$ ~

i j C

(u ,..\

y'$' r.b c, . ;. .

'v L\

(r,) 'Sr % '. iJ 1l:{ .g% ' ..f ; ~p3

-w . . .

so s y, 'y -

.\

,1,N,',y T. ~.

D0cy.*1 : lou . ',G-T)

%- yj 3

!etsruary '( , l'/O!

Pe u.ye the followin_ oneetn: g irt t he- followinn th rm t.n :

. ,Pm: 6-1, 6-2, 6-3, 6 h, 6-5, 6-6, Paces: 6-1, 6-2, 6-3, 6 4, 6-S , C-6, u-I, 6-3, o-9, 6-10, 6-11, 6-12, 6-17, 6-7, 6-8, c 's , 6-10, 6-11, G-w , 6-17, 6-18, 6-2;, and 6-22. 6-18, 6-J1, and G-22.

F1 cur e . - 6-1, 6-2, 6 h , and 6-5. icures: 6-1, 6-2, 6-h, and 6-5.

Pnses: 7 ') ana 7-10. Pa.nen : 7-9 and 7-10 F:rur+: 7-2. Figure: 7-2.

how B-;. B-2, 6-5, 8-6, 8-7, 8-8, Paces: 8-1, 6-2. 8-5, 8-6, 6-7, 8-6, e-9, 3-20, -; , and 8-12. 8-9, 8-10-, 6-11, and 6-12.

Fieuras: 8-1 and 8-2. Fieures: 8-1 and-8-2.

Pa m : 9 .3, 9 t. , 9-5, 9-6, 9-7, 9-8, Paees: 9-3, 9-4, 9-5, 9-6, 9-7, 9-8, 9-17, 9-18, 9-19, 9-20, 9-21, 9-22, 9-:7, 9-18, 9-19, 9-19a, 9-20, 9-21, 9-2 9-2 3, 9-24, 9-27, 9-28, 9-29, and 9-22a, 9-23, 9-24, 9-27, 9-28, 9-29, 9-30. and 9-30.

Figur es : 9-2, 9 4, 9-5, 9-6, 9-8, Fiaures: 9-2, 9 h , 9 ha, 9-5, 9-6, 9-8, 9-9, and 9-12. 9-9, and 9-12.

Gk Pages: 11-11,11-11a, 11-17, and 11-18. Pages: 11-11,11-11a, 11-17, and 11-16.

Pages: 1h-7, 1h-8, 1h-29, 1h-30, Paees: 14-7, 1h-8, 1h-29, 1h-30,-

14-37, 14-38, 14 bid, 14-42, lh h5, 14-37, 14-38, 1h 41d, lh h2, lh hs, and lh-h6. and ih h6.

Fig 2res: 11.-37 and ih-38. Figures: 1h-37 and ih-38.

Paces: 1A-9 and 1A-10. Pages: 1A-9 and 1A-10.

Pages: 5A-1 and SA-2. Pages: 5A-1 and 5A-2.

Supolement 1 Insert entire Supplement 1 in Volume k.

0029 b

a'

I Docket Nos. 50-302

' ' 50-303

1. fENDMENT NO. 1 i- v FLORIDA POWER CORPORATION

' Crystal River Plant Units 3 & L l

. The following sheets of the Florida Pcuer Corporation's Preliminary Safety Analysis lieport are to be deleted, and, where appropriate, revised sheets dated 1-15-68 should be inserted.

Renove tL fo11 cuing sheets: Insert the followine sheets:

Pages: Table of contents - iv and

~ Pages: Table of contents - iv and vi (one each in volunes 1, 2, 3, vi (one each in volunes 1, 2, 3, and h). and h).

1- 3, 1-h' , 1-13 , 1-14 , 1-2 5 , Pages: 1-3, 1 h, 1-13, 1-14, 1-25, I Pages:

1-26, 1-27, 1-28, 1-29, 1-30, 1-31, 1-26, 1-27, 1-28, 1-29, 1-30, 1-31, 1-32,1-33, 1-34, and 1-35 1-32, 1-33, 1-3h, and 1-35 1-2, 1-3, 1 h, 1-5, 1-6, Figures: 1-2, 1-3, 1-4, 1-5, 1-6, Figures: ~

1-7, 1 8, 1-9, 1-10, 1-11, and 1-14 1-7, 1-8, 1-9, 1-10, 1-11, and 1-14 i

2-3, 2-h, 2-5, 2-6, 2-7, 2-8, Pages: 2-3, 2 h , 2-5, 2-6, 2-7, 2-8, Paces: 2-11, and 2-12.

2-11, and 2-12. .

Pages: 3-vi, 3-3, 3-4, 3-5, 3-6, 3-9, Paces: 3-vi, 3-3, 3-4, 3-5, 3-6, 3-9, 3

3-10, 3-11, 3-12, 3-13, 3-1h, 3-31, 3-32, 3-10, 3-11, 3-12, 3-13, 3-1h , 3-31, 3-67, 3-68, 3-69, 3-70, 3-71, 3-72, 3-8T, 3-32, 3 67, 3-68, 3-69, 3-70, 3-71, 3-88, 3-89, 3-90, 3-91, 3-92, 3-95, 3-96, 3-72, 3-72a, 3-72b, 3-87, 3-88, 3-89, 3-97, 3-98, 3-103, and 3-104 3-90, 3-90a, 3-90b, 3-91, 3-92, 3-95, 3-96, 3-97, 3-98, 3-103, 3-103a, 3-103b , and 3-104.

~3-1, 3-2, 3-5, 3 6, 3-39, 3 h0, Figures: 3-1, 3-2, 3-5, 3-6, 3-39, Figures:

3-57, 3-58, 3-61, 3-62, 3-65, 3-66, 3-67, T 40, 3-57, 3-58, 3-61, 3-61a, 3-62, 3 65, 3-66, 3-67, and 3-68.

. and 3-68.

h-3, h-4, 4-15, 4-16, h-23, h-24, Pages: h-3, 4-4, h-15, h-16, h-23, '

1 Paces: 4-2h, h-25, h-26, h-27, h-28, 4-33, h-25, h-26, h-27, h-28, 4-33, h-3h, h-35, 4-34, h-35, and 4-36.

and h-36.

5-11, 5-111, 5-3, 5 h, 5-5, 5-6, Paces: 5-11, 5-111, 5-3, 5-4, 5 ha, Pages: 5-5, 5 6, 5-7, 5-7a, 5-8, 5-8a, 5-9, 5-7, 5-8, 5-9, 5-10, 5-11, 5-12, 5-13, 5-10, 5-11, 5-12, 5-13, 5-14, 5-15, 5-lh , 5-15, 5-16, 5-17, 5--18, and 5-20. 5-16, 5-16a, 5-17, 5-18, and 5-20.

I

~ 0030 0

A.i "- . . - . - _ . . . . _

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

Docket Nos. 50-302 1

50-303

• ice. move the following sheets: Insert the folicving cheets:

Figure

15-5. Figure: 5-5.

Pages: 6-11, 6-12, 6-13, 6-14,- 6-19, Pages: 6-11, 6-12, 6-13, c-14, 6-20, 6-21, and 6-22. 6-19, 6-20, 6-21, and 6-22.

Paceq: 7-5, 7-6, 7-7, 7-8, 7-17, Pacea: 72 5, 7-6, 7-7,'7-6, 7-17, 7-18, 7-19, 7-20, 7-23, 7-24, 7-27, 7-16, 7-19, 7-20, 7-23, 7-24, 7-27, 7-28, . 7-29, 7-30, 7-31, and 7-32. 7-28, 7-29, 7-30, 7-31, and 7-32.  !

F,igures : 7-2, 7-11, and 7-12. Figures : 7-2, 7-11, and 7-12.

Paces: ~

8-1, 6-11, 8-1, 8-2, 8-3, Paces: 8-1, 8-11, 6-1, 6-2, c-3, c-4, h4', 8-6, 8 i*,7 8-8, and 8-9 6 4, 8-5, 8-6, 8-7, 8-8, c-y, 6-10, 8-11, and 8-12.

' Figure: 8-1. . Figure: 8-1.

Paces: 9-1, 9-2, 9-7, 9-8, 9-9, 9-10, Pages: 9-1, 9-2, 9-7, 9-8, 9-9, 9-11, 9-12, 9-13, 9-14, 9-15, 9-16, 9-10, 9-11, 9-12, 9-13, 9-14, 9-15, 4 9-25, 9-26, 9-29, 9-30, and 9-37. 9-16, 9-25, 9-26, 9-29, 9-30, 9-37, and 9-37a.

Ficures: 9-3, 9-4, 9-8, eier-- Figures: 9-3, 9-4, 9-S, and 9-16 j Pages: 10-1, 10-2, 10-3, 10-4, and Pages: 10-1, 10-2, 10- 3, 10-4, and (4 10-5. 10-5 Figure: 10-1. Figure: 10-1.

i Pages: 11-3, 11 4, 11-5, 11-6, 11-7, Paces: 11-3, 11-h, 11-5, 11-6, 11-7,

11-8, 11-9, 11-10, 11-11, 11-12, 11-23, 11-6, 11-9, 11-10, 11-11,11-11a, i 11-24,'11-29, and 11-30. 11-12, 11-23, 11-24, 11-29, and 11-30. '

Figure,: 11-1. Figure: 11-1.

Paces: .12-1, 12-1, 12-2, 12-5, 12-6, Pages: 12-i, 12-1, 12-2, 12-5, 12-6, 12-7, and 12-8. 12-6a, 12-7, and 12-8.

Figure: 12-1. Ficure: 12-1.

s 0031 130 i

- + - -. .. = --.,.-.e ...-y-.... . .. . --. . . .. . . . .. ,,, , , , . ,

Ys s (s *. 2 4 *s4.p4s, J. .n . vy . . , , ' k , '*-'.

'g g dse 4 l'e. /Es'4A bE$ < .+ '"vd

~

f g g4pI ... _ , -

Locket Mos. 50-302 4

50-303 Remove the_ following sheets: Insert the following sheets:

O- Paces: 14-11, 14-v, 1h-3, 14-h, Pages: 1h-11, 14 , l'4-v(a ) , 14-3, 14-5, 1h-6, lh-7, 14-8, 14-9, 14-10, 1h-4, 14-5, 14-6, 1h-7, 14-6, 14-9, 1h-15, 1h-16,'14-17, 14-18, 14-21, 14-10, 1h-15, 14-16, 1h-17, 14-18, 14-22, 14-23, 14-24, 14-25, 14-26, 14-21, 14-22, 14-23, 14-24, 1h-25, 1h-27, 1h-28, 14-29, 14-30, 14-35, lh-26, 14-27, 14-28, 14-29,-14-30, 14-36, 14-39, 14 h0, 14-41, 1h-42, 1h-35, 14-36, 14-39,14-39a, 14 ho, 14-43. 14-kh, lb k9, 1h-50, 14-53, 14 h1, lk-kla, 14 hlb, 14-kle, ih kid, and 14-54 1h 42, lh-43, lb-kh, 14-49, 1h-50, 14-53, and 1h-54.

Ficures: lh-h2. Figures: lh-3ha, 14-3hb, lb-40a, 14 kla, lb-h2, 14 kha, lk-khb, 14 khe, ih khd, lk khe, and ih hkf.

Pages: 1A-1, 1/s-2, lA-3, and 1A 4. Pages: 1A-1, lA-2, lA-3, 1A-4, and 1A-ha.

Pages: Cover Sheet--Appendix 2C Pages: Cover Sheet--Appendix 2C and all sheets up to and including Cover Sheet- "2. Flood Studies for Crystal River Nuclear Power Plant."

NOTE Your revised Appendix 2C will now include two separate sections:

(1) " Plant Protection Against Hurri-cane Wave Action"...this is a new section included with this amendment and not foun. in the original P.S.A.R.

(2) " Flood Studies for Crystal River

~

Nuclear Power Plant"...the original Appendix 2C, which now becomes Section 2 of Appendix 2C.

Pages: 2G-1 to and including 2G-85 Pages: 2G-1 to and including 2G-119.

Figures: 2G-1, 2G-6, 2G-7, and 2G-8. Figures: 2G-1, 2G-6, 2L -7, and 2G-8.

Pages: 5A-1, 5A-2, 5A-3, and 5A-4. Pages: 5A-1, SA-2, SA-2a, 5A-3, and SA h.

Figure: Figure: SA-1.

Pages: 5B-1, 5B-2, 5B-3, 5B 4, and Pages: 5B-1, 5B-2, 5B-J, 5B h, 5B-5, 5B-5 5B-6, 5B-7, 5B-8, 5B-9, 5B 10 53-11,

4"

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p d TABLE OF CONTENTS Section P3ge, 1 INTRODUCTION AND SU'CIARY . . . . Volume 1 . . Tab 1 . . 1-1

1.1 INTRODUCTION

. . . . . . . . . . . . . . . . . 1-1 1.2 DESIGN HIGHLIGHTS . . . . . . . . . . . . . . . . 1-2 1.2.1 SITE CHARACTERISTICS . . . . . . . . . . . . . . 1-2 1.2.2 POWER LEVEL . . . . . . . . . . . . . . . . . 1-2 1.2.3 PEAK SPECIFIC POWER LEVEL . . . . . . . . . . . 1-2 1.2.4 REACTOR BUILDING SYSTE! . . . .- . . . . . . . . . 1-2 1.2.5 ENGINEERED SAFEGUARDS . . . . . . . . . . . . . . 1-3 1.2.6 ELECTRICAL SYSTE!S AND DIERGENCY POWER . . . . . . . . 1-3 1.2.7 ONCE-THROUGH STEAM GENERATORS . . . . . . . . . . . 1-4 1.3 TAEULAR CHARACTERISTICS . . . . . . . . . . . . . . 1h 1.4 PRINCIPAL DESIGN CRITERIA . . . . . . . . . . . . . 1-7 1.4.1 CRITERION 1 . . . . . . . . . . . . . . . . . 1-7 1.k.2 CRITERION 2 . . . . . . . . . . . . . . . . . 1-9 1.h.3 CRITERION 3 . . . . . . . . . . . . . . . . . 1-10 1.h.h CRITERION h . . . . . . . . . . . . . . . . . 1-10 1.h.5 CRITERION 5 . . . . . . . . . . . . . . . . . 1-11 1.h.6 CRITERION 6 . . . . . . . . . . . . . . . . . 1-11 1.h.7 CRITERION 7 . . . . . . . . . . . . . . . . . 1-12 1.h.8 CRITERION 8 . . . . . . . . . . . . . . . . . 1-13 1.h.9 CRITERION 9 . . . . . . . . . . 1-13

(~/'s .

A- 1.4.10 CRITERION 10 . . . . . . . . . . . . . . . . . 1-lk 1.h.11 CRITERION 11 . . . . . . . . . . . . . . . . 1-15 1.k.12 CRITERION 12 . . . . . . . . . . . . . . . . . 1-15 1.h.13 CR:TERION 13 . . . . . . . . . . . . . . . . 1-16 1.h.14 CRITERION lh . . . . . . . . . . . . . . . . . 1-17 1.h.15 CRITERION 15 . . . . . . . . . . . . . . . . 1-17 1.4.16 CRITERION 16 . . . . . . . . . . . . . . . . . 1-18 1.h.17 CRITERION 17 . . . . . . . . . . . . . . . . 1-19 1.h.18 CRITERION 18 . . . . . . . . . . . . . . . . . 1-20 1.h.19 CRITERION 19 . . . . . . . . . . . . . . . . 1-21 -

1.h.20 CRITERION 20 . . . . . . . . . . . . . . . . . 1-22 1.h.21 CRITERION 21 . . . . . . . . . . . . . . . . . 1-22 1.h.22 CRITERION 22 . . . . . . . . . . . . . . . . . 1-22 1.h.23 CRITERION 23 . . . . . . . . . . . . . . . . 1-23 1.h.2h CRITERION 2h . . . . . . . . . . . . . . . . 1-2h 1.h.25 CRITERION 25 . . . . . . . . . . . . . . . . 1-2k 1.h.26 CRITERION 26 . . . . . . . . . . . . . . . . . 1-25 1.h.27 CRITERION 27 . . . . . . . . . . . . . . . . 1-26 1.5 RESEARCH AND DEVELOPMENT REQUIRDIENTS . . . . . . . . . 1-26 1.5.1 CNCE-THROUGH STEAM GENERATOR TEST . . . . . . . . . . 1-26 1.5.2 CONTROL R0D DRIVE LINE TEST . . . . . . . . . . . . 1-26 ,

1.5.3 SELF-POWERED DETECTOR TESTS . . . . . . . . . . . . 1-27 1.5.h THERMAL AND HYDRAULIC PROGRAMS . . . . . . . . . . . 1-27 1.6 IDENTIFICATION OF AGENTS AND CONTRACTORS . . . . . . . . 1-27

1.7 CONCLUSION

S . . . . . . . . . . . . . . . . . 1-28

(.

i

{ Section .P,, age, 2 SITE AND ENVIRONMENT . . . . . Volume 1 . . . Tab 2 . . 2-1 2.1

SUMMARY

. . . . . . . . . . . . . . . . . . . 2-1 2.2 SITE AND ADJACENT AREAS . . . . . . . . . . . . . . 2-1 2.2.1 SITE LCCATION AND TOPCGRAPHY , . . . . . . . . . . 2-1 2.2.2 CITE CWNERSHIP . . . . . . . . . . . . . . . . 2-2 2.2.3 SITE ACTIVITIES . . . . . . . . . . . . . . . . 2-2 2.2.h F0FULATICN . . . . . . . . . . . . . . . . . 2-3 2.2 5 2-3 LAND USE . . . . . . . . . . . . . . . . . .

2.2.6 DAIRY ANIIIALS . . . . . . . . . . . . . . . 2-3 2.3 METEOROLOGY AND CLIMATOLOGY , . . . . . . . . . . . 2-3 2.3.1 SU:: MARY . . . . . . . . . . . . . . . . . . 2-3 2.3.2 DESCRIPTIVE METEOROLOGY . . . . . . . . . . . . . 2-h 2.3.3 ATMOSFHERIC DIFFUSION . . . . . . . . . . . . . . 2-7 2.3.h SITE METEOROLOGICAL PROGRAM . . . . . . . . . . . . 2-9 2.h HYDROLOGY . . . . . . . . . . . . . . . . . . 2-10 2.h.1 CHARACTERISTICS OF STREAMS IN VICINITY . . . . . . . . 2-10 2.h.2 FLOOD STUDIES AND HURRICANE EFFECTS . . . . . . . . . 2-11 2.h.3 DESIGN OF CIRCULATING WATER SYSTE4 . . . . . . . . . 2-11 2.h.h LIQUID WASTE DISCHARGES . . . . . . . . . . . . . 2-12 2.h.5 GROUNDWATER . . . . . . . . . . . . . . . . . 2-12 2.5 GEOLOGY . . . . . . . . . . . . . . . . . . . 2-12 2.6 SEISMOLOGY . . . . . . . . . . . . . . . . . . 2-13 2.6.1 SEISMICITY STUDY . 2-13

()

2.6.2 RESPONSE SPECTRA . . . . . . . . . . . . . . . 2-13 2.7 SITE ENVIRON'iENTAL BADI0 ACTIVITY PROGRAM , . . . . . . . 2-13

2.8 REFERENCES

. . . . . . . . . . . . . . . . . . 2-14 3 REACTOR . . . . . . . . . . Volume 1 . . . Tab 3 . . 3-1 3.1 DESIGN BASES . . . . . . . . . . . . . . . . . 3-1 3.1.1 PERFORMANCE OBJECTIVES . . . . . . . . . . . . . 3-1 3.1.2 LIMITS . . . . . . . . . . . . . . . . . . . 3-1 3.2 REACTOR DESIGN . . . . . . . . . . . . . . . . . 3-6

'i 2.1 GENERAL

SUMMARY

. . . . . . . . . . . . . . . . 3-6 3.2.2 NUCLEAR DESIGN AND EVALUATION . . . . . . . . . . . 3-7 3.2.3 THERMAL AND HYDRAULIC DESIGN AND EVALUATION . . . . . . 3-32 3.2.h MECHANICAL DESIGN LAYOUT . . . . . . . . . . . . . 3-68 3.3 TESTS AND INSPECTIONS . . . . . . . . . . . . . . 3-95 3.3.1 NUCLEAR TESTS ANI iSPECTION . . . . . . . . . . . 3-95 3.3.2 THERMAL AND HYDRAULIC TESTS AND INSPECTION . . . . . . . 3-95 3.3.3 FUEL ASSDIBLY, CONTROL R0D ASSEMBLY, AND CONTROL ROD DRIVE MECHANICAL TESTS AND INSPECTION . . . . . . . . . . 3-98 l

3.3.h INTERNALS TESTS AND INSPECTIONS . . . . . . . . . . 3-103 l

3.h . . . . . . . . . . . . . . . 3-10h REFEREECE3 . . .

4 REACTOR C001 ANT SYSTEM . . . . . Volume 1 . . . Tab h . . h-1 f

h.1 PERFORMANCE OBJECTIVES . . . . . . . . . . . . . . h-1 h-1

(~} h.1.2 DESIGN CEARACTERISTICS . . . . . . . . . . .

h-2 h.l.3 .

EXPECTED OPERATING CONDITIONS 00N .

(_/ . . .

11

4 Section Pm h REACTOR COOLANT SYSTE! ' CONTINUED) . Volume 1 . . . Tab h h.l.h SERVICE I" . . . . . . . . . . . . . . . . . h-3 h l.5 CODES A4.. CLASSIFICATIONS . . . . . . . . . . . . h-6

'h.2 SYSTE4 DESCRIPTION AND OPERATION . . . . . . . . . . . k-6 h.2.1 GENERAL DESCRIPTION . . . . . . . . . . . . . . h-6 h.2.2 MAJOR COMPONE:TS . . . . . . . . . . . . . . . 4-6 h.2.3 PRESSURE-RELIEVIli3 DEVICES . . . . . . . . . . . . h-12 h.2.4 ENVIR0!C! ENTAL PROTECTION . . . . . . . . . . . . . h-12 h.2.5 MATERIALS OF CONSTRUCTION . . . . . . . . . . . . h-12 h.2.6 MAXIMU'4 HEATING AND COOLING RATES . . . . . . . . . . h-lh 4.2.7 LEAK DETECTION . . . . . . . . . . . . . . . . 4-lh h.3 SYSTE4 DESIGN EVALUATION . . . . . . . . . . . . . h-16 k.3.1 SAFETY FACTORS . . . . . . . . . . . . . . . . 4-16 h.3.2 RELIANCE ON INTERCONNECTED SYSTE4S . . . . . . . . . h-23 h.3.3 SYSTDt INTEGRITY . . . . . . . . . . . . . . . 4-23 h.3.h PRESSURE RELIEF . . . . . . . . . . . . . . . . 4-23 h.3.5 REDUNDANCY . . . . . . . . . . . . . . . . . h-2h h.3.6 SAFETY ANALYSIS . . . . . . . . . . . . . . . . 4-2h h.3.7 OPERATIONAL LIMITS . . . . . . . . . . . . . . . h-24 h.h TESTS AND INSPECTIONS . . . . . . . . . . . . . . 4-25 h.h.1 COMPONENT IN-SERVICE INSPECTION . . . . . . . . . . h-25

, k.h.2 REACTOR COOLANT SYSTE4 TESTS AND INSPECTIONS . . . . . . 4-27

, h.h.3 MATERIAL IRRADIATION SURVEILLANCE . . . . . . . . . . k-28 i

h.5 REFERENCES . . . . . . . . . . . . . . . . . . h-30 5 CONTAINMENT SYST24 . . . . . . Volume 2 . . . Tab 5 . . 5-1 4

5.1 REACTOR BUILDING . . . . . . . . . . . . . . . . 5-1 5.1.1 DESIGN BASES . . . . . . . . . . . . . . . . . 5-1 5.1.2 STRUCTURE DESIGN . . . . . . . . . . . . . . 5-2 i 5.2 ISOLATION SYSTE! . . . . . . . . . . . . . . . . 5-9 5.2.1 DESIGN BASES . . . . . . . . . . . . . . . . . 5-9 l 5.2.2 SYSTE4 DESIGN . . . . . . . . . . . . . . . . 5-10' 5.3 VENTILATION SYSTE4 . . . . . . . . < . . . . . . 5-11 5.3.1 DESIGN BASES . . . . . . . . . . . . . . . . . 5-11 5.3.2 SYSTE4 DESIGN . . . . . . . . . . . . . . . . 5-12 5.h LEAKAGE MONITORING SYST24 . . . . . . . . . . . . . 5-13 5.5 SYSTEM DESIGN EVALUATION . . . . . . . . . . . . . 5-15 5.6 TESTS AND INSPECTION . . . . . . . . . . . . . . . 5-15 5.6.1 PREOPERATIONAL TESTING AND INSPECTION . . . . . . . . 5-15 5.6.2 POSTOPERATIONAL LEAK MONITORING . . . . . . . . . . 5-16 I

6 ENGINEERED SAFEGUARDS . . . . . Volume 2 . . . Tab 6 . . 6-1 6.1 atERGF.NCY INJECTION . . . . . . . . . . . . . . . 6-1 6.1.1 DESIGN BASES . . . . . . . . . . . . . . . . . 6-1 6.

1.2 DESCRIPTION

. . . . . . . . . . . . . . . . . 6-2 6.1.3 DESIGN EVALUATION . . . . . . . . . . . . . . . 6-3 6.1.h TESTS AND INSPECTIONS . . . . . . . . . . . . . . 6-6 6.2 . REACTOR B('TLDING ATMOSPHERE COOLING AND WASHING . . . . . . 6-13

. 6.2.1 l'ESIGN _dSES . . . . . . . . . . . . . . . . . 3 iii

Section' Page

(~%

(,'I 6 ENGINEERED _SAF_EGUARDS, (CONTINUED) . Volt =e 2 . . . Tab 6 6.2.2 DESCRIPTICH . . . . . . . . . . . . . . . . . 6-13 6.2.3 DESIGN EVALUATION . . . . . . . . . . . . . . . 6-lh 6.2.h TESTS AND INSPEC'.' IONS . . . . . . . . . . . . . . 6-19 6.3 ENGINEERED SAFEGUARES LEAKAGE A!!D RADIATION CONSIDEPATIONS . . . . . . . . . . . . . . . . 6-20 6.3.1 INTROLUCTION , . . . . . . . . . . . . . . . . 6-20 6.3.2 St!7ARY OF POSTACCIDENT RECIRCULATION AND LEAKAGE CONSIDERATIONS . . . . . . . . . . . . . . . 6-20 6.3.3 LEAKAGE ASSUMPTIONS . . . . . . . . . . . . . . 6-21 6.3.h DESIGN BASIS LEAKAGE . . . . . . . . . . . . . . 6-22 6.3.5 LEAKAGE ANALYSIS CONCLUSIONS . . . . . . . . . . . 6-22 7 CUSTRUMENTATION AIiD CONTROL . . . Volume 2 . . . Tab 7 . . 7-1 7.1 PROTECTION SYSTEMS . . . . . . . . . . . . . . . 7-1 7.1.1 DESIGN BASES . . . . . . . . . . . . . . . . . 7-1 7.1.2 SYSTEM DESIGN . . . . . . . . . . . . . . . . 7-5 7.1.3 SYSTEMS EVALUATION . . . . . . . . . . . . . . . 7-11 7.2 REGULATING SYSTEMS . . . . . . . . . . . . . . . 7-15 7.2.1 DESIGN BASES . . . . . . . . . . . . . . . . . 7-15 7 2.2 SYSTEM DESIGN . . . . . . . . . . . . . . . . 7-17 7.2.3 SYSTEM EVALUATION . . . . . . . . . . . . . . . 7-22 7.3

() 7 3.1 7.3.2 INSTRUMENTATION NUCLEAR INSTRUMENTATION .

NONNUCLEAR PROCESS INSTRUMENTATION 7-25 7-25 7-27 7.3.3 INCORE MONITORING SYSTEM . . . . . . . . . . . . . 7-28 7.h OPERATING CONTROL STATIONS . . . . . . . . . . . . 7-31 7.h.1 GENERAL LAYOUT . . . . . . . . . . . . . . . . 7-31 7.h.2 INFORMATION DISPLAY AND CONTROL FUNCTION . . . . . . . 7-31

)

7.h.3 SU! NARY OF ALARMS . . . . . . . . . . . . . . . 7-31 1 7.h.h COMMUNICATION . . . . . . . . . . . . . . . . 7-32 7.h.5 OCCUPANCY . . . . . . . . . . . . . . . . . . 7- 32 7.h.6 AUXILIARY CONTROL STATIONS . . . . . . . . . . . . 7-33 7.4.7 SAFETY FEATURES . . , . . . . . . . . . . . . . 7-33 8 ELECTRICALSYSIBjS Volume 2 . Tab 8 .

. . . . . . . . . 8-1 8.1 DESIGN RASES . . . . . . . . . . . . . . . . . 8-1 8.2 ELECTRICAL SYSTEM DESIGN

~ . . . . . . . . . . . . . 8-1 8.2.1 NETWORK INTERCONNECTIONS . . . . . . . . . . . . . 8-1 8.2.2 PLANT DISTRIBUTION SYSTEM . . . . . . . . . . . . 8-2 8.2 3 SOURCES OF AUXILIARY POWER . . .

8-6l1 8.3 TESTS AND INSPECTIONS . . . . . . . . . . . . . . 8-9 9 AUXILIARY AND D!ERGENCY SYSTD'S . . Volume 2 . . . Tab 9 . . 9-1 9.1 MAKEUP AND PURIFICATION SYSTEM . . . . . . . . . . . j-o O(I) a iv (Revised 1-15-68)

Section Pm 9 AUXILIARY AND E4ERGENCY SYSTE4S (CONTINUED) . Volume 2 . . Tab 9 9 1.1 DESIGN BASES . . . . . . . . . . . . . . . . 9-2 9 1.2 SYSTEM DESCRIPTION AND EVALUATION . . . . . . . . . . 9-3 9.2 CHEMICAL ADDITION AND SAMPLING SYSTEM . . . . . . . . . 9-9 9.2.1 . DESIGN BASES . . . . . . . . . . . . . . . . 9-9 9.2.2 SYSTE4 DESCRIPTION AND EVALUATION . . . . . . . . . . 9-10 9.3 COOLING WATER SYSTEMS . . . . . . . . . . . . . . 9-18 9.3.1 DESIGN BASES . . . . . . . . . . . . . . . . . 9-18 9 3.2 SYSTEM DESCRIPTION AND EVALUATION . . . . . . . . . 9-19 9.h SPENT FUEL COOLING SYSTE4 . . . . . . . . . . . . . 9-2h 9.4.1 DESIGN BASES . . . . . . . . . . . . . . . . 9-2h 9.k.2 SYSTEM DESCRIPTION AND EVALUATION . . . . . . . . . 9-2h 5.5 DECAY HEAT REMOVAL SYSTEM . . . . . . . . . . . . 9-27 9.5.1 DESIGN BASES . . . . . . . . . . . . . . . . 9-27 9.5 2 SYSTE4 DESCRIPTION AND EVALUATION . . . . . . . . . . 9-27 9.6 FUEL HANDLING SYSTEM . . . . . . . . . . . . . . 9-31 9.6.1 DESIGN BASES . . . . . . . . . . . . . . . . 9-31 9.6.2 SYSTEM DESCRIPTION AND EVALUATION . . . . . . . . . 9-32 9.7 PLANT VENTILATION SYSTEMS . . . . . . . . . . . . 9-37 9.7.1 DESIGN BASES . . . . . . . . . . . . . . . . 9-37 9.7.2 SYSTE4 DESCRIPTION AND EVALUATION . . . . . . . . . 9-37 10 STEAM AND POWER CONVERSION SYSTD1 . Volume 2 . . . Tab 10. . 10-1 10.1 DESIGN BASES . . . . . . . . . . . . . . . . . 10-1 10.1.1 0FJATING AND PERFORMANCE REQUIRE 4ENTS . . . . . . . . 10-1 10.1.2 ELECTRICAL SYSTEM CHARACTERISTICS . . . . . . . . . . 10-1 10.1.3 FUNCTIONAL LIMITATIONS . . . . . . . . . . . . . 10-1 10.1.4 SECONDARY FUNCTIONS . . . . . . . . . . . . . . 10-1 10.2 SYSTEM DESIGN AND OPERATION . . . . . . . . . . . . 10-2 10.2.1 SCHEMATIC FLOW DIAGRAM . . . . . . . . . . . . . 10-2 10.2.2 CODES AND STANDARDS . . . . . . . . . . . . . . 10-2 10.2.3 DESIGN FEATURES . . . . . . . . . . . . . . . . 10-3 10.2.4 SHIELDING , . . . . . . . . . . . . . . . . . 10-3 10.2.5 CORROSION PROTECTION . . . . . . . . . . . . . . 10-3 10.2.6 IMPURITIES CONTROL . . . . . . . . . . . . . . . 10-3 10.2.7 RADIOACTIVITY . . . . . . . . . . . . . . . . 10-3 10.3 SYSTEM ANALYSIS . . . . . . . . . . . . . . . . 10-3 10.3.1 TRIPS, AUTOMATIC CONTROL ACTIONS, AND ALARMS . . . . . . 10-3 10.3.2 TRANSIENT CONDITIONS . . . . . . . . . . . . . . 10-4 10.3.3 MALFUNCTIONS . . . . . . . . . . . . . . . . 10-5 10.3.4 OVERPRESSURE PROTECTION . . . . . . . . . . . . . 10-5 10.3.5 INTERACTIONS . . . . . . . . . . . . . . . . . 10-5 10.3.6 OPERATIONAL LIMITS . . . . . . . . . . . . . . . 10-5 10.h TESTS AND INSPECTIONS . . . . . . . . . . . . . . 10-5 h-V

Section Pace 11 RADI0 ACTIVE WASTES AND RADIATION PROTECTION . . . . . . . . Volume 3 . . . Tab 11. . 11-1 11.1 RADICACTIVE WASTES . . . . . . . . . . . . . . . 11-1 11.1.1 DESIGN BASES . . . . . . . . . . . . . . . . . 11-1 11.1.2 SYSTEM DESIGN . . . . . . . . . . . . . . . . 11-3 11.1.3 TESTS AND INSPECTIONS . . . . . . . . . . . . . . 11-12 11.2 RADIATION SHIELDING . . . . . . . . . . . . . . . 11-12 11.2.1 PRIMARY, SECONDARY, REACTOR BUILDING, AND AUXILIARY SHIELDING , . . . . . . . . . . . . . 11-12 11.2.2

• AREA RADIATICH MONITCRING SYSTEM . . . . . . . . . . 11-17 11.2.3 HEALTH PHYSICS . . . . . . . . . . . . . . . . 11-18 11.3 REFEPENCES . . . . . . . . . . . . . . . . . . 11-22 12 Volume 3 . Tab 12.

CONDUCT OF OPERATI.ONS

. . . . . . . . 12-1 12.1 ORGANIZATION AND RESPONSIBILITY , . . . . . . . . . . 12-1 12.1.1 FUNCTIONAL DESCRIPTION . . . . . . . . . . . . . 12-1 12.1.2 QUALIFICATIONS . . . . . . . . . . . . . . . . 12-2 12.1 3 ORGANIZATION DIAGRAM . . . . . . . . . . . . . . 12-2 12.2 TRAINING , . . . . . . . . . . . . . . . . . 12-2 12.2.1 STATION STAFF . . . . . . . . . . . . . . . . 12-2 12.2.2 REPLACEMENT PERSONNEL .

12.2.3

. . . . . . . . . . . . . 12-5 ON-THE-JOB TRAINING , . . . . . . . . . . . . . 12-6 12.2.4 EMERGENCY PLANS . . . . . . . . . . . . . . . 12-6 k#

1 12.2.5 PROCEDURES APPLICABLE TO ACCIDENTS INVOLVING RADIOACTIVE MATERIALS . . . . . . . . . . . . . 12-6 12.3 WRITTEN PROCEDURES 12.4 RECORDS .

. . . . . . . . . . . . . . . 12-7 4 12.5

. . . . . . . . . . . . . . . . . . 12-7 ADMINISTRATIVE CONTROL , . . . . . . . . . . . . . 12-7 13 INITIAL TESTS AND OPERATION Volume 3 .

. . . . . Tab 13. . 13-1 13.1 TESIS PRIOR TO REACTOR FUELING 13.2

. . . . . . . . . . . 13-1 INITIAL CRITICALITY . . . . . . . . . . . . . . . 13-1 13.3 POSTCRITICALITY TESTS . . . . . . . . . . . . . . 13-1 lh SAFETY ANALYSIS . . . . . . . Volume 3 . . . Tab 14. . 14-1 14.1 lb.l.1 CORE AND COOLANT BOUNDARY PROTECTION ANALYSIS

~~ . . . . . . 1h-1 ABN0EMALITIES . . . . . . . . . . . . . 14-1 lh.1.2 ANALYSIS OF EFFECTS AND CONSEQUENCES .

14.2

. . . . . . . . 1h-3 STANDBY SAFEGUARLS ANALYSIS . . . . . . . . . . . lh-20 14.2.1 SITUATIONS 'ANALYZED AND CAUSES . . . . . . . . .

. 1h-20 14.2.2 ACCIDENT ANALYSES

14.3 REFERENCES

. . . . . . . . . . . . . . . Ih-21

. . . . . . . . . . . . . . . . . . 1h-57 15- TECRNICAL SPECIFICATI0ES Volene 3 .

. . . . . . Tab 15. . 15-1 g 0039 U

vi (Revised 1-15-68)

.nz :. u,,: ne r:...s, . o:.:

i 's i

v/

Appendix

,. en.

. r--.c.::.,Lr C enr en ..r.rCn..Te-e

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2A FGFULATION AND LAND USE . . . . . . . . Appendices . . . . . Tab 2A 23 METEIROLOGI . . . . . . . . . . . . . App er.di c es . . . . Tab ?2

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12A 7LORIDA FUELIC SAFETY EMERGENCY FROCZDURES . . . . . . . . . . . . . . Appendices . . . . . Tab 12A Supplement 1 Response to DRL letter, 1-19-68. . . . . Volume 4 . . Supplement No. 1 2 Voluntary Response to AEC Oral Inquiry . Volume k . . Supplement No. 2 3 Voluntary Response to AEC oral Inquiry . Volume h . . Supplement No. 3 L Response to DRL letter, 3-13-69. . . . . Volume h . . Supplement No. 4 5 Response to DRL letter, 11-8-69. . . . . Volume 5 . . Supplement No. 5 6 Response to DRL letter, 11-8-69. . . . . Volume 5 . . Supplement No. 6 vii (3-2-70) h 1

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