ML19319C272

Chapter 4 of Davis-Besse PSAR, Rcs. Includes Revisions 1-8

ML19319C272
Person / Time
Site:	Davis Besse
Issue date:	08/01/1969
From:	TOLEDO EDISON CO.
To:
References
NUDOCS 8002110740
Download: ML19319C272 (65)

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PRELIMINARY SAFETY ANALYSIS REPORT .

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D-B TABLE OF CONTEITS VOLUME I Section Page 1 INTRODUCTION AND

SUMMARY

1-1

1.1 INTRODUCTION

1-1 1.2 DESIGN HIGHLIGHTS 1-2 1.3 TABULAR CHARACTERISTICS 1-T 1.h PRINCIPAL DESIGN CRITERIA 1_17 ,

15 RESEARCH AND DEVELOPMENT 1_h6 1.6 PROPOSED STATION DESIGN IN THE AREAS OF CONCERN IDENTIFIED IN A.C.R.S. LETI'FRS AS ASTERISKED ITEMS 1-53 1.7 THZ TOLEDO EDISON COMPANY COMPETENCE TO EUILD AND OPERATE DAVIS-BESSE NUCLEAR PO'ER STATION 1_56 1.8 IDENTIFICATION OF CONTRACTORS 1_56 19 CONCLUSIONS 1_56 2 SITE AND ENVIRONMENT 2_1 2.1

SUMMARY

21 2.2 SITE AND ADJACEiT AREAS 2-2 2.3 METECROLOGY 2_7 2.h HYDROLOGY 27 2.5 GEOLOGY 2-16 2.6 SEISMOLOGY 2-18 27 SU3S'URFACE CONDITIONS 2-20 2.8 SITE ENVIRONMENTAL RADIOACTIVITY PROGRAM 2-22 3 REACTOR 3-1 3.1 DESIGN BASES 3-1

.m 312 i

D-B TABLE OF CONTENTS (centd)

Section Page 3.2 EEACTOR DESIGN 3-6 3.3 TESTS AND UISPECTIONS ~

3.L REFERENCES 3-93 e

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D-B TABLE OF CONTENTS V0LLNE 2 Section Page h REACTOR COOLANT SYSTEM h-1 h.1 DESIGN BASES h-1 h.2 SYSTEM DESCRIPTION AND OPERATION h-5 h.3 SYSTEM DESIGN EVALUATION h-13 h.h TESTS AND INSPECTICUS h-21 h.5 QUALITY CONTROL h-2h h.6 REFEREICES h-29 5 CONTAIN!ENT 5-1 5.1 CONTAINMENT CONCEPT 5-1 5.2 CONTAINMENT SYSTEM STRUCTURAL DESIGN 5-3 5.) CONTAI UENT VESSEL ISCLATION SYSTEMS 5-28 5.h CONTAINMENT VESSEL COOLING AND VENTILATION SYSTEM 5-30 5.5 SHIELD BUILDING AND PENETRATION ROOM VENTILATION AND FILTRATION SYSTEM 5-32 5.6 LEAKAGE MONITORING SYSTEM 5-33 57 SYSTEM DESIGN EVALUATION 5-3h 5.8 TESTS AND INSPECTIONS 5-35 59 OTHER MAJOR STATION STRUCTURES 5 h0 5.10 FIFERENCES 5 h7 6 ENGINEERED SAFETY FEATURES 6-1 6.1 E ERGENCY CORE COOLING SYSTEM 6-2 6.2 CONTAINMENT ATMOSPHERE COOLING SYSTEMS 6-13 6.3 SHIELD BUILDING AND PENETRATICN ROOM VENTILATION AND ?ILTRATICH SYSTEM 6-20

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D-B TABLE OF CONTENTS (eontd) .

_Section Page 7 INSTRUMENTATIC+N AND CONTROL 7-1 7.1 REACTOR PROTECTION SYSTEM 7-1 72 SAFETY FEATURES ACTUATIcN SYSTEMS 7-13 7.3 REGULATING SYSTEMS 7-22 7.k NUCLEAR UNIT INSTRUMENTATICN 7-32 7.5 TURBINE CONTROL SYSTEMS 7-39 7.6 OPERATING CONTROL STATIONS 7 h0 sir J t,

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D-B TABLE OF CONTENTS VOLUME 3 Section M

8 ELECTRICAL SYSTD4S 8-1 8.1 DESIGN BASIS 8-1 8.2 ELECTRICAL SYSTEM DESIGN 8-1 8.3 TESTS AND INSPECTIONS 8-11 9 AUXILIARY AND EMERGENCY SYSTEMS 9-1 9.1 MAKEUP AND PURIFICATION SYSTEM 9-2 9.2 CHEMICAL ADDITION SYSTEM 9-9 9.3 COOLING WATER SYSTEMS 9-13 9.h SPENT FUEL COOLING SYSTEM 9-16 9.5 DECAY HEAT REMOVAL SYSTEM 9-18 9.6 FUEL HANDLING SYSTEM 9-21 97 SAMPLING SYSTEM 9-26 9.8 STATION VENTILATION SYSTEMS 9-27 99 INSTRUMENT AND SERVICE AIR SYSTEM 9-28 9.10 AUXILIARY FEEDWATER SYSTEM 9-30 9.11 FIRE PROTECTION SYSTEM 9-32 10 STEAM AND POWER CONVERSION SYSTE4 10-1 10.1 DESIGN BASIS 10-l*

10.2 SYSTEM DESCRIPTICN 10-1 10.3 SYSTEM RELIABILITY 10-h -

10.h TESTS AND INSPECTIONS 10 h 11 RADIOACTIVE WASTES AND RADIATION PROI'ECTION 11-1 11.1 RADIOACTIVE WASTES 11-1 N.*:

f v

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l D-B TABLE OF CONTDITS (contd)

Section Page 11.2 RADIATION PROTECTION 11-13

11.3 REFERENCES

11-22 12 CONDUCT OF OPERATIONS 12-1

12.1 INTRODUCTION

12-1 12.2 ORGANIZATION AND RESPONSIBILITY 12-1 12.3 PERSONNEL TRAINING 12-2 12.h WRITTEN PROCEDURE 12-6 12.5 RECORDS 12-6 12.6 ADMINISTRATIVE CONTROLS 12-6 13 INITIAL TESTS AND OPERATION iel 13.1 TESTS PRIOR TO REACTOR FUELING 13-1 13.2 INITIAL CRITICALITY 13-1 13.3 POSTCRITICALITY TESTS 13-1 14 SAFETY A'IALYSIS lh-1 1k.1 CORE AND COOLANT BOUNDARY PROTECTION ANALYSIS 1h-1 3h.2 STANDBY SAFETY FEATURES ANALYSIS lh-23

14.3 REFERENCES

lh-6h 15 TECHNICAL SPECIFICATIONS, 15-1 15.1 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SEI*'INGS 15-1 15.2 LIMITING CONDITIONS FOR OPERATION 15-2 15.3 SURVEILLANCE STANDARDS 15-7 <

15.h DESIGN FEATURES 15-9 15.5 ADMINISTRATIVE STANDARDS 15-10 317 mw d,., .

D-B

,'x LIST OF APPENDICES Appendix Volume 1A TECHNICAL QUALIFICATIONS 1 IB QUALITY ASSURANCE 1 1C PROJECT STAFF 1 2A RESTRICTED AREAS 1 2B METEOROLOGY l 2C GEOLOGY, SEISMOLOGY, SOIL AND FOUNDATION 1 -

DESIGN 2D LIMNCLOGY l 5A DESIGN BASES FOR STRUCTURES, SYSTEMS AND EQUIPMENT 2 53 DESCRIPTIONS OF LOAD FACTORS FOR SHIELD BUILDING

( AND CONTAINMENT INTERNAL STRUCTURE DESIGN 2 5C SPLICING REINFORCING BAR USING THE CADWELD PROCESS 2 5D JUSTIFICATION FOR YIELD REDUCTION FACTORS USED IN DETERMINING YIELD STRENGTH OF SHIELD BUILDING AND CONTAINMENT INTERNAL STRUCTURES 2 l

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^ TABLE OF CONTENTS s

s Section Page h REACTOR COOLA*iT SYSTEM h1 4.1 DESIGN BASES h_1 h.1.1 PERFORMANCE OBJECTIVES h_1 h.l.2 DESIGN CHARACTERISTICS h_1 h.l.2.1 Lesign Pressure h_1 h l.2.2 Lesien Te=rerature h_1 h.l.2.3 React?gn Leads h_1 -

4.1.2.h Seismic Loads and Loss-of-Coolant Leads h-2 4.1.2.5 Cyclic Loads h-3 h.l.2.6 Water Chemistry h-3 h.1.3 EXPECTED OPERATING CONDITIONS h-3 i

L 1.h SERVICE LIFE h-3 h.l.h.1 Material Radiatien Damage h_.

h l.h.2 Nuclear Unit Overational Thermal Cycles h_5 h.l.h.3 operatine Procedures h-5 2

h.l.h.h Cuality Manufacture h_6 h.l.5 CODES AND CLASSIFICATIONS h-6 h.2 SYSTEM DESCRIPTION AND OPERATION '

g4 L.2.1 GENERtL DESCRIPTION g_g h.2.2 MAJOR COMPONENTS h-7 h.2.2.1 Reactor Vessel h_7 h.2.2.2 Pressurizer h_7 4.2.2.3 . Steam Generator h_8

!. 2.2.h Reactor Coolant Pumps .

h_11 519 h-i Amendment No. 2

l D-B i

TABLE OF CONTENTS (Cont'd) l Section  !

Page h.2.2 5 Reactor Coolant Piping h-12 h.2.3 PRESSURE-RELIEVING DEVICES h-12 h.2.b ENVIRONMF.fiTAL PROTECTION h-12 h.2 5 MATERIALS OF CONSTRUCTION h-12 h.2.5.1 General h-12 h.2.5.2 .I

?ressure vessels  ;

h-13 4.2.6 MAXIMUM HEATING AND COOLING RATES k-lh "

4.2.7 l LEAK DETECTION h_1h h.3 I SYSTEM DESIGN EVALUATION h_1h h.3.1 SAFETY FACTORS h-lh 2 , h.3.1.1 Pressure Vessel (Including Steam Generator) Safety h_15 <

h.3.1.2 Piping h_19 h.3.1.3 Steam Generators h-20 h.3.2 FILIANCE ON INTERCONNECTED SYSTD!S h-22 h 3. 3 SYSTEM INTEGRITY h-22 l h.3.L PRESSURE PILIEF h-23 h.3 5 FIDUNDANCY h-23 h.3.6 SAFETY ANALYSIS h-23 h.3 7 OPERATIONAL LIMITS h-2h h.h TESTS AND INSPECTIONS I' h-25 {

h.h.1 COMPONENT IN-SERVICE INSPECTION '

l h-25 h.h 1.1 Reactor Vessel h-25 h.h.l.2 Pressurizer h-25 h.h.1.3 Steam Generator j2O h-25 h.h.l.h Reactor Coolan Pu=es k-26 Amendment No. 2 h-il

D-B p

TABLE OF CONTENTS (Cont'd) i .

- Section Page L.h l.5 Piping -

h-26 L.L.1.6 Essimilar Metal and Retresentative Welds 4-26 h.h.1.7 Instection Schedule h-26 L.4.2 REACTOR COOLANT SYSTEM TESTS h-26 h.h.2.1 Reactor Coolant System Precritical and Hot Functional Test h-26 L.h.2.2 Pressurizer Precritical Overational Test h-27 h.h.2.3 Relief Valves Test h-27 h.h.2.h Unit Power Startup Test h-27

/ 4.h.2.5 Unit Power Heat Balance g,37 h.h.2.6 Unit Po.rer Shutdevn Test h-27 h.h,3 MATERIAL IRRADIATION SURVEILLANCE h-27 2 h.5 QUALITY CONTROL h-27 h.5.1 GENERAL h-27 h.5.1.1 Dimensional Inspection h-27 k.5.1.2 Nondestructive Testing h-28 L.5.1.3 Welding and Heat Treating Processes h-29 h.5.1.4 Material Identification h-30 h.5.2 REACTOR VESSEL h-30 h.5.3 STEAM GENERATOR h-31 h.5.h PRESSURIZER h-31 h.5.5 REACTOR COOLANT PIPING h-31 h.5.6 REACTOR COOLANT PUMP CASINGS s h-32 h.5.7 VALVES

. h-32 h.6 REFERENCES h-32 v

h-lii Amendment No. 2

D-B LI&r OF TABLES /

(At Bear of Fectien)

Table No. Title b-1 Tabulation of Reactor Coolant System Pressure Settings 4-2 Reactor Ccolant Quality h-3 Reactor Vessel Design Data Lh Pressuriser Design Data k-5 Stes:n Generator Design Data h-6 Steam Generator Feedvater Qua.'ity 4-7 Reactor Coolant Pu=p Design Data t

h-8 Reactor Coolant Piping Design Data h-9 Transient Cycles h-10 Reactor Coolant System Codes and Classifications b-11 Materials of Construction h-12 References for. Figure h-k - Increase in Transition Temperature Due to Irradiation Effects for A302B Steel 322 h-iv

D-B LIST OF FIGURES (At Rear of Section)

Figure No.

h-1 Reactor Coolant System h-2 Reactor Coolant System Arrangement-Elevation h-3 Reactor Coolant System Arrangement-Plan hh Nil-Ductility Transition Temperature Increase Versus Integrated Neutron Exposure for A3023 Steel h-5 Reactor vessel h-o Pressuriser ~

h-7 Steam Generator h-8 9 team Generator Heating Regions h-9 Steam Generator Heating Surface and Downcomer Level Versus Power h-10 Steam Generator Temperatures h-11 Reacter Coolant Pump h-12 Predicted NDTT Shift Versus Reactor Vessel Irradiation 323 h-v

D-B O h REACTOR COOLANT SYSTEM

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The reactor coolant system consists of the reactor vessel, coolant pumps, steam generators, pressurizer, and interconnecting piping. The functional relation-ship between coolant system components is shown in Figure h-l. The coolant system physical arrangement is shown in Figures h-2 and L-3 To assist in the review of the system drawings, a standard set of sy=bols and abbreviations has been used and is summarized in Figure 9-1.

h.1 DESIGN BASES 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 re-flecter, and as a solvent for the soluble boron utilized in chemical shim re-activity control.

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

'4hile performing this function, the system serves the safeguards objective of minimizing the release to the reactor building of fission products that es-eape the primary barrier, the fuel cladding.

l2 h.l.2 DESIGN CEARACTERISTICS h.1.2.1 Desi n Pressure The reactor coolant system design, operating, and control set point pressures are listed in Table h-l. The design pressure allows for operating transient

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pressure changes. The selected design =argin considers core thermal lag, cool-ant transport times and pressure drops, instrumentation and control response characteristics, and system relief valve characteristics. The design pressures and data for the respective system components are listed in Tables h-3, h h, L-5, k-7, and h-c.

h.1.2.2 Desizn Te=cerature The design temperature for each component is selected above the maximum antici-pated coolant temperature in that ec=ponent under all nonnal and transient load conditions. The design and operating temperatures of the respective system com-ponents are listed in Tables h-3, h-4, h 3, h-7, and h-8.

L.l.2.3 Reacticn *. cads s

All ec=ponents in the reactor coolant system are supported and interconnected so that piping reaction forces result in combined mechanical and thermal stresses in equipment no::les and structural valls within established code limits. Equip-ment supports are designed to transmit piping rupture reaction loads to the massive shielding and foundata.. structures.

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h-1 Anendment No. 2 J

D-B h.l.2.h Seismic Loads and Loss-of-Coolant Loads Reactor coolant system components are designated as Class 1 equipment and are designed to maintain their functional integrity during an earthquake and loss-of-coolant accident. Design is in accordance with the design bases shown below.

The loading combinations and corresponding design stress criteria for internals, vessels, supports, and piping are given in this section. A discussion of each of the esses of loading combinations follows: Table 5A-2 in Appendix 5A pre-sents the stress limits for the loading combinations discussed as applicable to all Class 1 equipment.

In Cases II and III, secondary stresses are neglected since they are self-limit-ing. Design stress limits in most cases are in the plastic region, and local yielding would occur. Thus, the conditions that caused the stresses are as-sumed to have been satisfied.

h.l.2.h.1 Seismic Loads Case I - Design Loads Plus Design Earthouake (Maximum Probable) Loads For this combination, the reactor must be capable of continued operation; there-fore, all the above components excluding piping are designed to the allowable 3

stress limits of Section III of the ASME Code for Reactor Vessels. The primary piping is designed according to the requirements of ANSI B31.7'. The S valces 3

for all components, excluding bolting, are those specified in Table N-421 of the ASME Code. The S= value for bolts are those specified in Table N-h22 of the ASME Code.

Case II - Design Loads Plus Maximum Hypothetical Earthquake (Maximum Possible) Loads In establishing stress levels for this case, a "no-loss-of-function" criterion applies, and higher stress values than in Case I can be allowed. The condition is defined as a faulted condition according to Paragraph N-bl2 (t) (h) of Sec-tion III. The S values m are those specified in Paragraph N-h17.ll of the ASME Code or Paragraph h.1.2.h.2a.

4.1.2.k.2 Loss-of-Coolant Loads A loss-of-coolant accident coincident with a seismic occurrence will be analyzed to assure the ability to initiate and maintain a reactor shutdown and emergency core cooling. The following loading case is considered:

Case III - Design Loads Plus Maximum Hypothetical Earthquake (Maximum Possible) Loads Plus Pipe Runture Loads The design stress limits will be based on either of the following: "

a. Two-thirds of the ultimate strength of the material. The design allowable stress of Case III loads is given in Figure 3-1 for 30h ,.

stainless steel. This curve is used for all reactor internals , in- J 2 I) cluding bolts. It is based on adjusting the ultimate strength curves s published by U.S. Steel to minimum ultimate strength values by using )

the ratio of ultimate strength given by Table N-h21 of Section III ,

' Amendment No. 3 h-2

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D-B of the ASME Code at room temperature to the rocm te=perature strength

'% given by U.S. Steel.

b. Stress limits in accordance with Paragraph N h17.ll of Section III 6

for faulted conditions with the exception of 90% of the plastic instability loads .

h.l.2.5 Cyclic Loads All components in the reactor coolant system are designed to withstand the effects of cyclic loads due to reactor system te=perature mad pressure changes.

These cyclic loads are introduced by normal unit lead transients, reactor trip, and startup and shutdown operation. Design cycles are shown in Table h-9.

During unit startup and shutdevn, the rates of temperature and pressure changes are limited.

h.l.2.6 Water Chemistry The water che=istry is selected to provide the necessary boron content for reactivity centrol and to minimize corrosion of reactor coolant system sur-faces. Water coolant quality specifications are given in Table h-2. The reactor coolant chemistry is discussed in further detail in 9.2.

h.l.3 EXPECTED OPERATING CONDITIONS Throughout the lead range frca 15 to 100 per cent power, the reactor coolant system is operated at a constant average te=perature. Reactor coolant system pressure is centrolled to provide sufficient overpressure to maintain adequate core subcooling.

The minimum operating pressure is established from core thermal analysis. This analysis is based upon the maximum expected inlet and outlet te=peratures, the maximum reactor power, the minimus DNBR required (including instrumentation errors and the reactor centrol system deadband), and a core flow distributica factor. The maximum cperating pressure is established on the basis of ASME Code relief valve characteristics and the margins required fo. 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 pressuriner heater acticn to compensate for low pressure. Normal operational lifetime transient cycles are discussed in detail in h.1.h.2.

h.l.h SERVICE LIFE The service life of reactor coolant system pressure ce=ponents depends upon the end-of-life material radiation damage, nuclesr unit operational ther'nal cycles, quality manufacturing standards, environmental protection, and adherence to established operating procedures.

s h.1.h.1 Material Radiation Damage MRie reactor vessel is the only reactor coolant system component exposed to a significant level of neutron irradiation mad is therefore the only such com-ponent subject to material radiation damage. The design value for the fast neutrcn nyt with energies greater than 1.0 MeV at the inner surface of the re-actor vessel is 3.0 x 1019 n/cm2 . The correspcnding calculated fast neutron h-3 Amendment 6 b

D-B flux et tho v;ssal vall is 2.h x 10 n/cm -soc at 2,788 MWt. This calculated value includes a lifetime average axial peaking factor of 1.3 and an azimuthal peaking factor of 1.29. For LO years at 80% load this corresponds to an nyt of 2.h x 1019 n/cm2 ,

The attenuation of the neutron flux from the core to the reactor vessel is ccm-puted using the NRN(1) program. This is a one-dimensional multigroup removal-liffusicn program in slab, cylindrical, or spherical geometry. This code uses the method in which the uncollided and strongly forward-scattered neutrons (re-moval groups) are computed by integration of an energy dependent attenuation kernel over the source volume. Scattering of neutrons out of the removal groups forms a source term for the multigroup diffusion calculations. Neutron slow-ing-down is handled by elastic and nonelastic scattering =atrixes for both the removal and diffusion groups.

Neutron fluxes at the vessel vall are co=puted with the core represented as a slab source equal in thickness to the equivalent core diameter. A lifetime average power distribution through the thickness of the core was determined .

from calculated power profiles over several core cycles and at various times during each cycle. The neutron energy spectrum was represented by 26 energy .

groups with lh of these groups covering the range above 1.0 MeV. '

Local flux peaking on the vessel vall due to fuel assemblies extending beyond the equivalent core diameter (acimuthal peaking) is determined with PDQ 5t2),

a two-dimensional diffusion program. The lifetime-average axial flux peak at the vessel is the same as that in the two outer rows of fuel assemblies.

Calculations with the NRN code were compared with data from various exp ments including measurements on the R2- eactor at Studsvik Research Center, on the LIDO pool reactor at Harve11, and on a shielding mockup of B&W's 800 MWe reactor vessel and internals design at the B&W Critical Experiment Labora-tory (CEL).(5) At the R2-0 reactor, measurements were made throu~h about 3 feet of water with thresholds detectors which included ll5In (n,n ) ll5 min (1.5 steV), 32s (n,p) 32P (3 MeV), and 27A1.(n,a) 2hNa (6 MeV). Energies shown are threshold energies for the reactions. In the LIDO pool, thermal flux measure-ments were made through laminations of iron and water over a penetration distance of about h feet. In the experiment at B&W's CEL, sulphur foil data was taken at points covering the distance between the core and the reactor vessel.

In all cases, and over the entire penetraticn distances, the calculations were either in agreement with the data, or predicted higher flux and activation levels. It is thus concluded that the NRN code provides a reliable method for

'he calculation of vessel nyt.

For the design neutron exposure, the predicted Nil-Ductility Transition Temper-ature (NDTT) sM t is 250 F based on the " Maximum Curve for 550 F Data" shown in Figure 4-5. $ Based on a specified initial NDTT of 10 F, this shift would result in a predicted maximus NDTT of 260 F.

The " Trend Curve for 550 F Data," as shown in Figure h-h, represents irradiated <

material test results and was compiled from the reference documents listed in Table b-12.

h-h

D-B h.l.h.2 I;uclear Unit Crerational Thermal Cycles

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To establish the service life of the reactor coolant system ccmponents as re-quired by the ASMI III for Class "A" vessels, the nuclear unit operating cen-ditions that involve the cyclic application of leads and thermal conditions have been established for the LO-year design life.

The number of thermal and loading cycles to be used for design purposes are listed in Table L-9, " Transient Cycles." The estimated actual cycles based on a review of existing nuclear stations operations are also provided in Table L-9 Table L-10 lists the design codes and classifications for reactor cool-ant system ccmpcnents. The effect of individual transients and the sum of inese transients ara evalusted to determine the fatigue usage factor during the detail design and stress analysis effort. As specified in ASMI III Para-graph L15.2 (d)(6), the cumulative fatigue usage factor vill be less than 1.0 for the design :ycles listed in Table h-9 The transient cy: lea listel in Table L-9 are conservative and complete in that tney include all significant acdes of ncr:a1 and emergency Operation. The es-timated frequenc;, 'rsses for the design transient cycles are listed in Table k-9 A larger number cf cycles of smaller magnitudes than those described can be tol-ersted.

A heatup and ecoldevn rate of 100 F/ hour is used in the analysis of Transients i and 2 in Table k-9 A ramp loading and rs=p unicading transient is defined as a change in pcVer level frcn 15 to 100 to 15 per cent of rated pcVer at a rate of change of 5%/

min between 15 and 100 per cent, both up and dcun. A step leading transient is an instantanecas pcVer increase or decrease of 10 per cent of rated power.

A step load reduction to auxiliary lead is an instantaneous reduction in elec-trical lead frca 1G0 to 5 per cent of rated load.

The miscellaneous transiente (Item d) listed in Table L-9 include the initial hy'drotests, plus an allevance for ft.ture hydrotests in the event that reactor coolant syste= =cdifications or repairs may be required. Subsequent to a nor-

=al refueling operation only the reactor vessel closure seals are hydrotested for pressure integrity; therefore, reactor acolant system hydrotests before startup are not included.

h.l.L.3 Crerating Procedures The reacter coolant system pressure retaining actponents are designed using the transition temperature method of cinimizing the possibility of brittle fracture of the vessel caterials. The varicus cc=binations of stresses are evaluated and enployed to detarmine the system operating procedures.

The basic determination of vessel operation frc cold startup and shutdevn to full pressure and tenperature operaticn is performed in acccrdance with a " Frac-ture Analysis Diagram" as published by Fellini and Pu:ak.(I)

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D-B At temperatures below the Nil.-Ductility Transition Temperature (NDTT) and the Design Transition Temperature (I7fT), which is equal to NDTT + 60 F, the pres-sure vessels vill be operated so that the stress levels will be restricted to h

a value that vill prevent brittle failure. These levela are:

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

b. From the temperature of DTT 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 stresses are held within the referenced stress limits (a through c above),

brittle fracture vill not occur. This statement is based on data reported by Robertson(8) and Kihara and Masubichi(9) in, published literature. It can be shown that stress limits can be controlled by imposing operating procedures .

that control pressure and temperature during heatup and cooldown.18) This pro-cedure vill insure that the stress levels do not exceed those specified in a through e above.

h.l.k.h Quality Manufacture The quality control program for the reactor coolant system is as outlined in h.5 This program vill be monitored as described in Appendix 13.

h.l.5 CODES AND CLASSIFICATIONS All pressure-containing components of the reactor coolant system ar Jigned, fabricated, inspected, and tested te, applicable codes as listed in , , h-10.

The pump casings will be designed to meet and will be certified by tne pump vendor to have been fabricated to meet the requirements of ASME III, but they will not be code stamped.

h.2 SYSTEM DESCRIPTION AND OPERATION h.2.1 GENERAL DESCRIPTION The reactor coolant system consists of the reactor vessel, two vertical once-through steam generators, four shaft-sealed coolant circulating pumps, an elec-

.trically heated pressurizer, and interconnecting piping. The system is arranged as two heat tranr.pcrt loops, each with two circulating pumps and one steam gen-erator. Reactor system design data are listed in T'.ble 4-3, h-h, h-5, h-7, and h-8, and a system schematic diagram is shown in Figure b-1. Elevation and plan views of the arrangement of the major co=ponents are shown in Figures h-2 e and h-3. .

329 J u.s

D-B h.2.2 MAJOR COMPONENTS t

h.2.2.1 Reactor Vessel The reactor vessel consists of a cylindrical shell, a spherically dished bottom head, and a ring flange to which a removable reactor closure head is bolted.

The reactor closure head is a spherically dished head welded to a ring flange.

The vessel has six major no::les for reactor coolant flow, 69 control rod drive no::les mounted on the reactor closure head, and two core flooding no::les -

all located above the core. The reactor vessel vill be vented through the con-trol rod drives. The vessel closure seal is formed by two concentric 0-rings with provisions between them for leakage collection. The reactor vessel, no:-

le design, and seals incorporate the extensive design and fabrication experi-ence accumulated by B&W. Fifty-seven in-core instrumentation no::les are lo-cated on the lower head.

The reactor closure head and the reactor vessel flange are joined by sixty 6-1/2 in. diameter studs. Two metallic 0-rings seal the reactor vessel when the recctor closure head is bolted in place. Leakoff and test taps are pro-vided in the annulus between the two 0-rings to dispose of leakage and to hydro-test the vessel closure seal after refueling.

The vessel straight cylindrical shell sections are insulated with metallic re-flective-type insulation. Conventional mass 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, plenum assembly, fuel as-semblies, control rod assemblies, and incore instrumentatien.

The reactor vessel shell material is protected against fast neutron flux and ga==a heating effects by a series of water annuli and the thermal shield located between the core and vessel vall. This protection is further described in 3.2.h.1, k.l.h, and h.3.1.

Guide lugs velded to the reactor vessel inside vall limit reactor internals and core vertical drop to 1/h in. or less and prevent rotation about the verti-cal axis in the unlikely event of a major core barrel or core support shield failure.

The react >r vessel general arrangement is shown in Figure h-5, and the general

• arrangement of the reactor vessel and internals is shown in Figures 3-h5 and 3-h6. Reacter vessel design data are listed in Table h-3

• .2.2.2 Presauriner _

The general arrangement of the reactor coolant system pressurizer is shown in Figure h-6, and the design character stics are tabulated in Table h k. The l

B0 h-7

D-B electrically heated pressurizer establishes and maintains the reactor coolant pressure within prescribed limits and provides a surge chamber and a vater re- Qs serve to accommodate reactor coolant volume changes during operation.

The pressurizer is a vertical cylindrical vessel connected to the reactor outlet piping by the surge piping. The pressurizer vessel is protected from ther=al effects by a thermal sleeve on the surge line and by a distribution baffle located on the surge pipe inside the vessel.

Two ASME Code relief valves are mounted on the top of the pressurizer and func-tion to relieve any system overpressure. Each valve has one half the required relieving capacity. The capacity of +hese valves is discussed in !. 3.k. An additional pilot-operated relief valve is provided to limit the lifting fre-quency of the code relief valves. The relief valves discharge to a pressurizer quench tank located within the reactor building. The quench tank has a stored water supply to condense the steam. A rupture disc protect the tank against overpressure should a pressurizer relief valve fail to reseat.

The pressurizer contains replaceable electric heaters in its lover section and a water spray nozzle in its upper section to maintain the steam and water at the saturation temperature corresponding to the desired reactor coolant system pres-sure. 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 normal operating pressure. During insurges, as pressure in the reactor system increases, steam is condensed by a water spray from a reactor inlet line, thus reducing pressure. Spray flow and heaters m are controlled by the pressurizer pressure controller.

]

Instrumentation for the pressurizer is discussed in 7.h.2.

h.2.2.3 Steam Generator The general arrangement of the steam generators is shown in Figure h-7, and de-sign data are tabulated in Table 4-5.

The steam generator is a veitical, straight-tube-and-shell heat exchanger and produces superheated steam at constant pressure over the power range. Reactor coolant flovs' downward through the tubes, and steam is generated on the shell side. The high pressure parts of th9 unit are the hemispherical heads, the tubesheets, and the straight In0cnel P) 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 boundaric of the steam-producing section of the vessel. Within the shell, the tube bundle is surrounded by a baffle, which is divided into two sections. The upper part of the annulus between the shell and baffle is the superheater outlet, and the lower part is the feedvater inlet-heating zone. Vents, drains, instrumentation _

v' <

P )Inconel is a trade name of an alloy manufactured by the International Nickel Company. It also has substantial common usage as a generic description of a Ni-Fe-Cr allc,y conforming the ASTM Specification SB-163. It is in the latter '

context that reference is made here.

h-8 bb*

D-B nozzles, and inspection openings are provided on the shell side of the unit.

.N, The reactor coolant side has manways on both heads, and a drain nos le for the

\ botto= head. Venting of the reactor coolant side of the unit is accomplished by a vent connection on the reactor coolant inlet pipe to each unit. The unit is supported by a shirt attached to the bottom head which rests on a sliding support and provides the required freedom of movement to accommodate reactor coolant system thermal expansion.

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

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

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

. cold feedvater.

The saturated water in the downec=er 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 fon=ed by the down-comer, the boiling sections, and the bypass steam flov to the feed-water heating region. With lov (less than 1 ft/sec) saturated water velocities entering the generating section, the secondary side pres-sure drops in the boiling section are negligible. The majority of the pressure drop is due to the static head of the mixture. Conse-quently, the downco=er 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 stea=-vater 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 DNB, i.e. , departure from nucleate boiling, is reached, and then film boiling and superheating occurs.

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

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

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

h-9 9 k

D-B Shown on Figure h-9 is a plot of heating surface and downec=er level versus load. As shown, the downcomer water. level is proportional to steam flov from 15 - 100 per cent load. A ccnstant minimum level is held below 15 per cent I) load. The a=ount of surface (of length) of the nucleate boiling section and the film boiling section is proportional to lead. The surface available for superheating varies inversely with load, i.e., as load decreases the superheat section gains from the nucleate and film boiling regions.

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

The simple concept with ideal counterflow conditions results in highly stable flow characteristics en both the reactor coolant and secondary sides. The hot reactor coolant fluid is cooled uniformly as it flows downward. The secondary side = ass flow is lcw, 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 ficw of water and steam.

A plot of reactor ecolant and steam temperatures versus reactor power is shown 2l in Figure 7 h. As shown, both steam pressure and average reactor coolant tem-perature are held constant over the lead range from 15 to 100 per cent rated power. Constant steam pressure is obtained by a variable two-phase boiling length (see Figure h-9) and by the regulation of feed flow to obtain proper steam generator secondary = ass inventory. In addition to average reactor cool-ant temperature, reactor coolant flow is also held constant. The difference between reactor coolant inlet and. outlet temperatures increases proportionately as load is increased. Saturation pressure and temperature are constant, re-

)/

sulting in a variable superheater outlet temperature.

Figure h-10, a plot of te=perature versus tube length, shows the temperature differences between shell and tube throughout the steam generator at rated lead.

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

The shell temperature is controlled by the use of direct contact steam that heats the feedwater to saturation, and the shell is bathed with saturated water from feedwater inlet to the lower tubesheet.

In the superheater section, the tuie vall temperature approaches the reactor coolant fluid te=perature since the steam film heat transfer coefficient is 7l 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 no::le, further reducing temperature differentials between tubes and shell.

The steam generater design and stress analysis will be performed in accordance with *be requirements of the ASME III as described in h.3.1.1.

Table h-6 provides specifications of steam generator feedwater quality.

~

bb l

Amendment No. 7 k-10 ,

l rM

D-B h.2.2.h Reactor Coolant Pu=es The general arrange =ent of a reactor coolant pu=p is shavn in Figure h-ll, and the pump design data are tabulated in Table 4-7. The reactor coolant pumps are vertical, single-stage, single-suction, . shaft-sealed units . The pu=p casing consists of a bottem suctica inlet passage which delivers the fluid to the pu=p impeller, a multivaned diffuser and a collecting scroll which directs the fluid out through a horizontal discharge nozzle.

Each pump has a separate, single-speed, top-mounted electric drive rotor, which is connected to the pu=p by a removable shaft coupling. A motor stand is used as the transition piece between the motor and the pu=p, which in conjunction with the removable shaft coupling, allows pu=p seal replace =ent without re= oval of the drive motor.

Ghaft sealing is accc=plished in the upper part of the pu=p housing using a removable seal cartridge which centains three mechanical seals in series. A drain chamber is mounted above the top seal to collect the last seals lubricating 2' and leakage water. Water for lubricating and cooling the seals is injected belev the bottom seal at a pressure approxi=ately 50 psi above reactor system pressure.

Part of the seal flow passes into the pu=p casing through a radial restriction bus hing. The re=ainder f1cvs up through the seal cartridge where its pressure and flow is equally divided by the three seals and is then returned to the seal water supply system.

Low pressure ecoling water is also furnished to the pu=p as a backup to the seal

, injection water system. The cooling water flows to a heat exchanger mounted integral to the pu=p. The heat exchanger is sized to, upon loss of seal inje'c-tion water, provide enough cooling capacity to prevent excessive heating of the

=echanical seals to occur. The upper drain chamber, which features an overflow drain to a closed disposal system, further prevents leakage to the reactor building if excessive deterioration of the =echanical seal perfor=ance should occur.

A water-lubricated, self-aligning, radial hydrostatic bearing is located in the pump housing. Two oil-lubricated, radial bearings and a Kingsbury type, double-l2 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 lubrication is maintained by internal pumping provisions in the motor, or by an external system which is re- f6 quired for " hydraulic-jacking" of the bearing surfaces for startup.

An antirotatien device vill be furnished with each pump =otor to prchibit reverse rotation of the pu=p.

Thrust bearing, vibration, and seal performance tests will be made in a closed loop on the first pump at rated rpeed with the pu=p end at rated temperature and pressure. Sufficient testing vill be done on subsequent units to substantiate that they conform to the initial test pump characteristics.

04 M h-ll Amendment No. 6

D-B h.2. 5 Reactor coolant Piping O

The general arrangement of the reactor coolant system piping is shown in Figures 4-2 and h-3. Piping design data are presented in Table h 8. In addition to the pressurizer surge piping connection, the piping is equipped with velded connec-tions for pressure taps, te=perature elements, vents, drains, decay heat removal, and emergency core cooling high-pressure injection water. Ther=al sleeves are provided in piping connections when the ther=al transient design analysis indi-cates that this protection is required.

h.2.3 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 located on the top head of the pressurizer. The relief valves discharge into the pressurizer quench tank which condenses and collects the effluent. The schematic ,

arrangement of the relief devices is shown in Figure h-l. Since all sources of heat in the system, i.e., core, pressurizer heaters, and reactor coolant pu=ps, are interconnected by the reactor coolant piping with no intervening isolation valves, all relief protection can conveniently be located on the pressurizer.

k.2.k ENVIRON! ENTAL PROTECTION The reactor coolant system is surrounded by concrete shield walls as described in Section 5.

Lateral bracin6 vill be provided near thc steam generator upper tubesheet eleva- ~

t'on to resist lateral loads, including those resulting from seismic forces, pipe rupture, thermal expansion, etc. Additional bracing is provided at a lovar ')

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

h.2.5 MATERIALS OF CONSTRUCTION h.2.5.1 General Each of the naterials used in the reactor coolant system has been selected f 3r the expected environmental and service conditions. The major component mtcrials are listed in Table h-11.

All reactor coolant system =aterials normally exposed to the coolant are corrosion-resistant materials consisting of 30h or 316 SS, Inconel (Ni-Cr-Fe),17 k PH (H1100), or veld deposits with corrosion-resistant properties equivalent to or better than those of 30h SS. These materials were chosen for specific purposes ct various locations within the system because of their superior co=patibility with the reactor coolant.

Periodic analyses of the coolant chemical composition will be performed to monitor the adherence of the system to the reactor coolant water qualit; listed in Table '

h-2. Maintenance of the water quality to minimize corrosion is performed by tne chemical addition and sampling systems which are described in detail in 9.2 and 9.7.

The feedvater quality entering the steam generator will be held within the limits n(75 h h-12 1

D-B

, , k.2.5.2 Pressure Vessels The reactor vessel plate material opposite the core is purchased to obtain a Charpy V-notch test result of 30 ft-lb or greater at a corresponding nominal low-Nil-Ductility Transition Temperature (NDTT), and the material vill be tested to determine the actual NDTT value. In addition, this plate vill oe 100 per cent volu=etrically inspected by ultrasonic test using both normal and shear wave.

The reactor vessel material is heat-treated specifically to obtain good notch-ductility qualities. The stress limits established for the reactor vessel are dependent upon the temperature at which the stresses are applied. As a result of fast neutron absorption in the region of the core, the =aterial ductility will change. The effect is an increase in the NDTT. The predicted maximum end-of-life NDTT value of the reactor vessel opposite the core, based on an initial value of 10 F, vould be 260 F. The predicted neutron exposure and NDTT shift are discussed in h.l.k.

The unirradiated or initial NDTT of pressure vessel base plate material is mea-

~

sured by the Charpy V-notch impact test (Type A) given in ASTM E23. Using the Charpy V-notch test, the UDTT 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 Section III Table N-332 specifies an energy value of 30 ft-lb. A curve of the temperature versus the energy absorbed in breaking the specimen is plotted. To obtain this curve at least two specimens are tested at a mini-mum of five different temperatures. Available data indicate NDTT differences as great as 40 F between curves plotted through the minimum and avera6e values respectively. The intersection of the energy versus temperature curve with the 30 ft-lb ordinate is designated as the NDTT. The determination of NDTT from the average curve is considered representative of the material and is consis-tent with procedures specified in ASIM E23. The material for these tests is treated P- the methods outlined in ASME III Paragraph N-333. The test coupons are take.1 at a distance 'of T/h (1/h of the plate thickness) from the quenched surfaces and at a distance of T from the quenched edges. These tests are per-

~

for=ed by the material supplier or B&W in accordance with ASME Section III, Paragraphs N-313 and N-330.

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

k.2.6 E*XIMUM HEAIING AND COOLING RATES The normal reactor coolant system operating cycles are given in Table h-9 and described in h.1.h.2. The normal system heating and cooling rate is 100 F/hr.

The fastest cooldown rates resulting from the break of a main steam line are discussed in lb.l.2 9 00suguP5 I

'I /, *

, t h-13 tilu

D-B h.2.7 LEAK DETECTION To =inimite leakage frc= the reactor coolant system all ecmponents are inter-connected by an all-velded piping system. Some of the components have access openings of a flanged-gasketed design. The largest of these is the reactor closure head, which has a double metal 0-ring seal with provisions for dis-posing of leakage between the 0-rings.

With regard to the ree.ctor vessel, the probability of a leak occurring is con-sidered to be remote en the basis of reactor vessel design, fabrication, test, inspection, and cperation at temperatures above the material NDTT as described in h.3.1. Reactor closure head leakage vill be zero from the annulus between the metallic C-ring seals during vessel steady-state and virtually all transient operating conditions. Only in the event of a rapid transient operation, such as an e=ergency cooldown, would there be some leakage past the innermost 0-ring seal. A stress analysis en a similar vessel design indicates this leak rate vould be approximately 10 cc/ min through the seal monitoring taps to a drain, and no leakage vill occur past the outer 0-ring seal. The exact nature of this transient condition and the resulting small leak rate vill be determined by a detailed stress analysis.

In the unlikely event that significant leakage should occur frem the system 2 into the containment during reactor operation, the leakage vill be detected by one or more of the following methods:

a. Total reactor coolant system leakage rate can be determined by ec=-

paring the instrument-recorded indications of reactor coolant aver-age temperature and pressurizer and makeup tank water level over a time interval. The pressurizer is maintained at a constant level, and any coolant leakage results in a decrease of makeup tank water level. The makeup tank capacity is 31 gallons per inch of tank height, and each graduation on the level recorder represents 2 inches of tank height.

b. Control room instrumentation vill indicate an increase in containment 2 activity,

c. Changes in containment humidity and sump level instrumentation 2

indicators are indications of leakage into the containment.

h.3 SYSTEM DESIGN EVALUATION L.3.1 SAFETY FACTORS The reactor coolant system is designed, fabricated, and erected in accordance with proven and recognized design codes and quality standards applicable for ~

the specific compcnent function er classification. These components are de-signed for a pressure of 2,500 psig at a nominal te=perature of 650 F. The corresponding nominal operating press'tre 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 h-10.

Amendment No. 2 h-lh

7 D-B Aside from the safety factors introduced by code requirements and quality con-trol programs, as described in the following paragraphs, the reactor coolant I

system functional safety factors are discussed in Sections 3 and lb.

4.3.1.1 Pressure Vessel (Including Steam Generator) Safety The safety of the nuclear reactor vessel and other reactor coolant system pres-sure vessels, including the reactor coolant pump casing, is dependent upon four major factors': (a) design and stress analysis, (b) quality control, (c) proper operation, and (d) additional safety factors. The special care and detail used in implementing these factors in pressure vessel =anufacture are briefly de-scribed as follows:

h.3.1.1.1 Design and Stress Analysis These pressure vessels are designed to the requiremen's t 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 ves-sels. It is a ec=prehensive code based on the most applicable stress theory. '

It requires a stress analysis of the entire vessel under both steady-state and transient operations. The result is a ec=plete evaluation of both primary and secondary stresses, and the fatigue life of the entire vessel. This is a con-trast with previous codes which basically established a vessel thickness dur-ing steady-state operations only. In establishing the fatigue life. of these pressure vessels, using the design cycles frem Table b-9, the fatigue evalua-tion curves of ASME III are employed.

Since ASME III requires a ec=plete 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 solutiens. The 3&W Cc=pany has confirmed 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 those predicted by the theory. E&W has also conducted laboratory deflection studies of thick shell and ring ec=binations to define the accuracy of the theory, and is using ec=puter progra=s developed on the basis of this test data.

The analytical procedure considers all process operation conditions. A detailed design and analysis of every part of each Section III vessel is prepared as fol-lows:

a. The vessel size and configuration are set co 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 no::les, etc., required by ASME III.

b. The vessel pressure and te=perature design transients given in Table h-9 are employed in the determination of the preasure loading and temperature gradient and their variations with time throughout the vessel. The resulting combinations of pressure loading and thermal g stresses are calculated. Ccmputer programs are used in this develop-J ment.

h-15g

r D-B

c. The stresses through the vessel are evaluated using as criterir the allowable stresses per ASME III. This code gives safe stress level limits fdr all the types of applied stress. These are membrane strecs . Q' )

(to insure adequate tensile strength of the vessel), membrane plus primary bending stress (to insure a distortion-free vessel), secondary stress (to insure a vessel that will not progressively deform under cyclic loading), and peak stresses (to insure a vessel of maximum fatigue life). A design report is prepared and submitted to the jurisdictional authorities and regulatory agencies, i.e. , state, insurance, etc. This report defines in suffi-cient detail the design basis, loading conditions, etc., and will summarize the conclusions to permit independent checking by interested parties. In evaluating the integrity of the reactor coolant system, it is important to consider that the size of the defect in a system component that can conceivably contribute to the rupture of a vessel depends on the size and orientation of the defect, the magnitude of the stress field, and tem These major param-eters have been correlated by Pellini and Puzak(I)perature.who have prepared a " Fracture Analysis Diagram" which is the basis of vessel operation from cold startup and shutdown to full pressure and temperature operation. The diagram predicts that, for a given level of stress, larger flaw sizes will be required for fracture initiation above the NDTT temperature. For example, at stresses in the order of 3/h yield strength, a flaw in the order of 8 to 10 in. may be sufficient to initiate fracture at tempera ^ures below the NDTT tem-perature. However, at NDTT +30 F, a flaw of 1-1/2 times this size may be re- -~ quired for initiation of fracture, while at a temperature of NDTT +60 F, brittle )

                                                                                             ~

fracture is not possible under elastic stresses because brittle fracture propa-gation does not take place at this temperature. Fractures above this tempera-ture are of the predominantly ductile type and are dependent upon the member net section area and section modulus as they establish the applied stress. Fracture Mode Evaluation An analysis has been made to demonstrate that the reactor vessel can accommo-date without failure the rapid temperature change associated with the postulated operation of the emer$ency core cooling system (ECCS) at end of vessel design life. A summary of the evaluation follows: The state of stress in the reactor vessel during the loss-of-coolant accident was evaluated for an initial vessel temperature of 603 F. The inside of the vessel vall is rapidly subjected to 90 F injection water of the maximum flow rate obtainable. The results of this analysis show that the integrity of the vessel is not violated. The assumed modes of failure are ductile yielding and brittle fracture, which includes the nil-ductility approach and the fracture mechanics approach. The i modes of failure are considered separately as follows:

a. Ductile Yielding The criterion for this mode of failure is that there shall be no gross yfelding across the vessel wall using the minimum specified yield y ,

h-16 j3h

D-B , strength in the ASME Code, Section III. The analysis considered the maximum combined thermal and pressure stresses through the vessel wall thickness as a function of time during the safety 17.jection. Com-parison of calculated stresses to the material yiela stress indicated that local yielding may occur in the inner 8.0 per cent of the vessel wall thickness.

b. Brittle Fracture Because the reactor vessel vall in the core region is subjected to neutron flux resulting in embrittlement of the steel, this area was analyzed from both a nil-ductility approach and a fracture mechanics approach. The results of the two methods of analysis compare favorably and show that pressure vessel integrity is maintained.

The criterion used in the nil-ductility approach is that a crack cannot propagate beyond any point where the applied stress is below the thres-hold stress for crack initiation (5 - 8 ksi) or when the stress is ccm- . pressive. This approach involves making the very conservative assump-tion that all of the vessel material could propagate a crack by a low energy absorption or cleavage mode. End-of-life conditions were as-sumed. The crack arrest temperature through the thickness of the wall was developed on a stress-temperature coordinate system. The actual quench-induced, stress-temperature condition through the thick-ness of the wall at several times during the quench was developed and plotted. The maximum depth at which the material in the vessel vall would be in tension or at which the stress in the material would be in excess of the threshold stress for crack initiation (5 - 8 ksi) was determined by comparison of the plots. The comparison showed that a crack could propagate only through the inner 35 per cent of the wall thickness if a crack initiation threshold of 5 - 8 ksi is applicable. The foregoing method of analysis is essentially a stress analysis ap-proach which assumes the worst conceivable material properties and a flaw size large enough to initiate a crack. Actually, the outer 83 per cent of the vessel vall is at a temperature above the DTT (NDTT + 60 F) when credit is taken for the neutron shielding, and for the original DTT profile through the wall thickness. The analysis is conservative in that it does not deny that cracks can be initiated, and in that it assumed a crack from 1 to 2-ft long to exist.in the vessel wall at the time of the accident. Therefore, it can be con-cluded that, if a crack were present in the worst location and orienta-tion (such as a circumferentially oriented crack on the inside of the vessel wall), it could not propagate through the vessel vall. A fracture mechanics analysis, was conducted which assumed a continuous surface flaw to exist on the inside surface of the vessel wall. The c criterion used for the analysis is that a crack cannot propagate when the stress intensity at the tip of the crack is below the critical crack stress intensity factor (KIC)* Babcock & Wilcox Topical Report No. 10018; ".4nalysis of the Structural Integrity of'the Reactor Vessel Subjected to a Thermal Shock" provides N c u_,, cf 340

D-B the details of the analysis. This report includes an evaluation considering the Irvin fracture mechanics method and performs a sen- n . sitivity analysis of_the effect of varying the conservatism of several

 ,             major parameters on the result.

Reactor Vessel Closure Studs The reactor closure head is attached to the reactor vessel with sixty, 6-1/2 in. diameter studs. The stud material is A-540, Grade B23 (ASME III, Case 1335) , which has a minimum yield strength of 130,000 psi. The studs, when tightened for operating conditions, will have a tensile stress of approximately 30,000 psi. Thus, at operating conditicns (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 remaining studs reach the ulti-  !

mate tensile strength.

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

The stress' analysis of the studs vill include a fatigue evaluation. It is not expected that fatigue evaluation vill yield a significantly high usage factor for the LO-year design life. m Control Rod Drive Service Structure "' The control rod drive service structure is designed to support the control rod drives to assure no loss of function in the event of a combined loss of cool-ant accident and maximum hypoti.etical earthquake. Requirements for rigidity, imposed on the structure to avoid adversely affecting the natural frequency of vibration of the vessel and internals, as well as space requirements for service routing, result-in stress levels considerably lover than design limits. The structure is more than adequate to perform its required function. 4.3.1.1.2 operation As previously mentioned in h.l.h, pressure vessel service life is dependent on adherence to established operation 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 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 Committee. The stress limits established  ; for the pressure vessels . are dependent upon the te=perature at which the stresses are applied. As a result of fast neutron absorption in the region of the core, the reactor vessel material ductility will change. The effect is an increase in the Nil-Ductility _ Transition Temperature (NDTT). The determination of the predicted  ; I

                              -             - h-18        -

p t  ;

D-B

 .Ay NDTT shift is described in 4.1.h.l. This NDTT shift is factored into the unit startup and shutdown procedures so that full cperating pressure is not attained until the reactor vessel temperature is above the design transition te=perature (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 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, h.l.h.2, and h.l.h.3.

4.3.1.1.3 Additional Reactor Vessel Safety Factors Additional methods and procedures used in reactor vessel design, not previously mentioned in h.3.1.1 above but which are considered ccuservative and provide an additional margin of safety, are as follows: (DELETED) 13

a. Use of minimum specified yield strength of the material insteed of the actual values,

b. Neglecting the increase in yield strength resulting from irradiation effects.

c. The design shift in NDTT as given in h.l.h.1 is based on maximum pre-dicted flux levels at the reactor vessel inside vall surface, whereas the bulk of the reactor vessel material vill experience a considerably lover exposure to radiation and consequently a lower change in NDTT over the life of the vessel.

d. Results from the method of neutron flux calculations, as described in 4.1.h.1, have been checked closely with experiment, and include an appropriate factor to account for azimuthal and undertainty tri-tien in the nyt at the reactor vessel vall. The conservative assuc; _vus, model description, and comparison of calculational code results with experiment are discussed in detail in h.1.h l.

e. The effect of thermal shock resulting frcm actuation of the emergency core cooling system on reactor vessel integrity har been analyzed.

The foregoing descriptions of vessel design, fabrication, and cperating pro-cedures present a base for continued confidence in the integrity of the reacter coolant system components. Reference 10 and B&W experience support the conclu-sion that reactor vessel rupture is incredible. h.3.1.2 Piping Design, stress analyses, quality centrol and operating limits for the reactor coolant piping vill provide a level of system integrity equivalent to that , achieved for the reactor coolant system pressure vessels. i # e 1 3R h-19 Amendment No. 3

D-B Total stresses resulting from thermal expansion and pressure and mechanical and ceismic loadings are all considered in the design of the reactor coolant piping. } The total stresses that can be expected in the piping are within the maximum code allowables. Connections, when required, are equipped with thermal sleeves to limit stresses from thermal shock to acceptable values. All materials and fabrication proce ures will meet the requirements of the specified code. All material will be ultrasonically inspected. All interior surfaces of the inter-connecting piping are clad with stainless steel to eliminate corrosion problems and to reduce coolant contamination. Pressure welds and cladding will be in-spected by nondestructive techniques including radiographic, dye penetrant, or magnetic particle examinations as appropriate. h.3.1.3 Steam Generators Research and Develorment The steam generator research and development program has been ecmpleted. The , program included the following elements of testing:

a. The steady state and transient operation tests have confirmed the analytically predicted performance characteristics of the steam gen-erator and have provided the data for the control system.

b. Feedwater spray nozzle tests have demonstrated that the design will satisfactorily heat the feedwater.

c. Tube leak simulation tests have demonstrated that a leak in one tube j will not propagate by causing a failure in adjacent tubes.

d. Mechanical tests have demonstrated that the tubes can withstand, without failure, the mechanical loads they may experience either during normal operation or accident conditions.

e. Vibration testing demonstrated that the unit contained no undesirable resonance characteristics.

f. Tests to simulate a steam line failure or reactor coolant system fail-ure have demonstrated the integrity of the steam generator under con-ditions of rapid depressurization and large temperature differentials between the tubes and the shelt of the unit.

3 The work is reported in a proprietary Babcock & Wilcox Topical Report BAW 10002,

 ^
   "Research and Development Report for the Once Through Steam Generator."

Stress Evaluation Because the steam generator is of a straight tube-straight shell design and be-cause of a minor difference in the coefficient of thermal expansion between Inconel and carbon steel, there exists a structural limitation on the mean temperature difference between the tubes and the shell. During normal opera-tion of the steam generator, the tube mean temperature should not be more than 32 F higher than the shell mean temperature. The maximum calculated mean tube i to shell AT at normal operating conditions poses no problems to the structural s'

                                                                    'A d b .7) e h-20 L

D-B integrity of the reactor coolant boundary. The effect of loss of reactor cool-ant would impose tensile stresses on the tubes and cause slight yielding across the tubes. Suen'a condition would introduce a small pecmanent deformation in the tubes but would in no way violate the boundary integrity. The rupture of a cecondary pipe would cause the tubes to become warmer than the shell and may

 .cause tube deformation. Blowdown tests simulating secondary side blowdown on a 37-tube model boiler, show that although slight buckling in the tubes occurred, there was no loss of reactor coolant.

Calculations confirm that the steam generator tube sheet will withstand the loading resulting from a loss of coolant accident. The basis for this analysis is a hypothetical rupture of a reactor coolant pipe resulting in a maximum de-sign pressure differential from the secondary side of 1,050 psi. Under these conditions there is no rupture of the primary to secondary boundary (tubes and tube sheet). The maximum primary membrane plus primary bending stress in the tube sheet under . these conditions is 15,900 psi across the center ligaments which is well below the ASME Section III allowable limit of h0,000 psi at 650 F. Under the condition postulated, the stresses in the primary head show only the effect of its role as a structural restraint on the tube sheet. The stress intensity at the juncture of the spherical head with the tube sheet is 1h,970 psi which is well below the allowable stress limit. It can therefore be concluded that no damage vill occur to the tube sheet or the primary head as a result of this postulated accident. In regard to tube integrity under loss of reactor coolant, actual pressure tests of 5/8 in. OD/0.034 inch wall Inconel tubing show collapse under an external pressure of h,950 psi. This is a factor of safety of 4.7 against collapse under the 1,050 psig accidental application of external pressure to the tubes. The rupture of a secondary pipe has been assumed to impose a maximum design pressure differential of 2,500 psi across the tubes and tube sheet from the primary side. The criterion for this accident permits no violation of the reactor coolant boundary (primary head, tube sheet, and tubes). T o meet this criterion, the stress limits delineated in the ASME Pressure Ves-sel Code, Section III, Paragraph N-714.2 for hydrotest limitations are applicable for the aforementioned abnormal operating circumstan'e. The referenced section-states that the primary membrane stresses in the tube sheet ligaments, averaged across the ligament and through the tube sheet thickness, do not exceed 90% of the material yield stress at the operating temperature; in addition the pri-mary membrane plus primary bending stress in the tube sheet ligaments, averaged across the ligament vidth at the tube sheet surface location giving a maximum stress, does not exceed 135% of the material yield stress at the operating tem-perature. e An examination of stresses under these conditions show that for the case of a

 .2,500 psig design pressure differential, the stresses are within acceptable limits. These stresses together with the corresponding stress limits are given in Table k-13                                                                              ;

i The bssic design criterion for the tubes assumes a pressure differential of ' l 2,500 psts in accordance with Section III. Therefore, the secondary pre sure CI' 344  ! 4-21 l

D-B loss accident condition imposes no extraordi'tary stress on the tubes beyond 7q'

       'that normally expected and concidered in Section III requirements.                   '
                                                                                               ,/

The superimposed effect of secondary side pressure loss and maximum hypothetical , earthquake has been considered. For this condition, the criterion is that there be no violation of the primary to secondary boundary (tube and tube sheet). For i the case of the tube sheet, the maximum hypothetical earthquake loading vill contribute an equivalent statt.: pressure loading over the tube sheet of less than 5 psi (for vertical shock). The effect of fluid dynamic forces on the steam generator internals under secon-dary steam break accident conditions has been simulated in a 37 tube laboratory boiler. Results of the tests show that reactor coolant boundary integrity is maintained under the most severt mode of secondary blowdown. The ratio of allowable stresses (based on an allowable membrane stress of 0 9 of the nominal yield stress of the material) to the computed stresses for a design pressure differential of 2,500 psig are summarized in Table h-lh.  : h.3.2 RELIANCE ON INTERCONNECTED SYSTEMS The principal heat removal systems which are interconnected with the reactor coolant system are the steam and feedvater systems and the decay heat removal system. The reactor coolant system is c'.ependent upon the steam generators, and the ste m, feedwater, and condensate s'; stems for-decay heat removal from noual operating conditions to a reactor coolant temperature of approximately 280 F. All vital-active components in these systems are duplicated for reliability purposes. Flow diagrams of the steam and feedwater systems are 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 atmosphere. Two turbine-

driven auxiliary feed pumps will supply water to the steam generators.

The decay heat removal system is used to remove decay heat when the thermal )I driving head of the reactor coolant system is no longer adequate to generate steam. This system is completely described in 9.5 The heat received by this system is rejected to the service water system which also contains sufficient redundancy to insure adequate operation. A schematic diagrcm of the decay heat removal system is presented in Figure 9-7 h.3.3 SYSTEM INTEGRITY The integrity of the reactor coolant system is insured by proper materials ' selection, 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 (NDTT) to eliminate the possibility of propagating-type failures. The react.or coolant system is designed in 3:cordance with ASME pressure vessel

  .3   and ANSI power pipisg codes as given in T ule 4-10. Relief valves on the pres-           T surizer are sized to prevent system prest ce from exceeding the design point          ,j by more than 10_per cent.

Amendment No. 3 >

h-22

D-B Ar a further assurance of system integrity, all components in the system will be hydrotested at 3,125 psig before initial operation. The largest and most frequently used opening in the reactor coolant system, i.e., the reactor clo-sure head, contains provisions for separate hydrostatic pressurization between the 0-ring type gaskets. h.3.h PRESSURE RELIEF The reactor coolant system is protected against overpressure by relief valves located on the top of the pressuri;er. The capacity of the pressurizer relief valves is determined from considerations of (a) the reactor protection system, (b) pipe pressure drop (static and dy-namic) between the point of highest pressure in the reactor coolant system and the pressurizer, and (c) accident or transient conditions that may potentially cause overpressure. Preliminary analysis indicates that the hypothetical case of withdrawal of a regulating control rod assembly bank from a relatively low power provides the - basis for establishing pressurizer relief valve capacity. The accident is terminated by ugh-pressure reactor trip with resulting turbine trip. This ac-cident condition produces a power mismatch between the reactor coolant system and steam system larger than that caused by a turbine trip without immediate reactor trip, or by a partial load rejection from full load. The ASME III required relief valve capacity as determined on the basis of the accident described above is 600,]00 lb/h. Two 300,000 lb/h valves are in-stalled. A pilot-operated relief valve, capable of 100,000 lb/h relief, is provided to limit the lifting frequency cf the code relief valves. h.3.5 REDUNDANCY The reactor coolant system contains two steam generators and four reactor cool-ant pumps. For added reliability, power to the pumps is available from two separate busses as shown in Section 8. Separate core flooding nozzles are provided on opposite sides of the reactor vessel to insure core reflooding water in the event of a single core flooding nozzle failure. Reflooding water is available from either the core flooding tanks or the decay heat pumps which provide lev-pressure coolant injection as an engineered safety feature. Thehigh-pressureinjectionpipesareconnectedi 1 to the reactor coolant system on each of the four coolant inlet nines. h.3.6 SAFETY ANALYSIS The components of the reactor coolant system are interconnected by an all-welded piping system. Since the reactor inlet aad outlet nozzles are located above the core, there is never any danger of the reactor coolant uncovering " the core when any other system component is drained for inspection or repair.

 ~ Complete strety analyses are included in Lection lb.
  %)
                                                            ~

3'46 h-23 (#

D-B h.3 7 OPERATIONAL LIMITS , Reactor coolant system heatup and cooldown rates are described in detail in h.1.4 and h.2.6. The component stress limitations dictated by material IUIT considerations are described in h.l.k 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 582 F at rated power. In this mode of operation, the reactor vessel outlet temperature is 608 F. Additional temperature variations at various power levels are shown on Figure 7-5 Reactor trip signals will be fed to the reactor protection system as a result of high reactor coolant temperature, high pressure, low pressure, and low flow, i.e., flux-flow comparison and pump monitor signals. By relating lov flow to the reactor power, operation at partial power is feasible with less than four reactor coolant pumps operating. The reactor coolant system operating limits are as follows: - Performance Vs Pu=ps In-Service Reactor Coolant 2 Pumps 2 Pumps Pumps Ocerating h Pu=ps 3 Pu=ps (One in Each Loop) (1 Loop) Maximum Reactor Power, 100 75 h5 40 3

   ". of Rated                                                                         J Reactor Coolant Flov,       100       Th              48               h6
   ". of Rated Reactor operating limits under natural circulation conditions are discussed in 1h.1.2.6.3. The bases for the selection of operational limits are discussed further in 7.1.2.h.

The reactor coolant system is designed for continued operation with one per cent of the fuel rods in the failed condition. The radioactivity content of the coolant is based on long-term saturation activities with one per cent failed fuel (Table 11-3).

                                                                                    'o 3_23 347

t D-B h.h TESTS AND INSPECfIONS k.h.1 COMPONENT IN-SERVICE INSPECTION Consideration has been given to the inspectability of the reactor coolant sys-tem in the design of components, in the equipment layout, and in the support structures to permit access for the purposes of inspection. Access for inspec-tion is defined as access for visual examination by direct or remote means and/ or by contacting vessel surfaces during nuclear unit shutdown. h.h.l.1 Reactor Vessel Access for inspection of the reactor vessel vill be as follows:

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

b. External surfaces and velds on the closure head can be inspected vis-ually or by contact following removal of the insulation. Internal -

surfaces of the closure head can be examined visually by remote means.

c. Inner surfaces of the vessel outlet nozzles can be inspected visually by remote means during refueling periods. The complete internal sur-face can be :hspected by remote visual means following removal of the reactor core and vessel internals.

d. External surfaces of the vessel nozzle to piping velds can be in-spected by remote visual means following removal of shielding and vessel insulation.

e. The external surface of the reactor vessel can be inspected during the reactor lifetime if it should become necessary. An annulus has been provided bet' teen the reactor vessel and the primary shield to accccmodate inspection requirementc.

h.h.1.2 Pressurizer The external surface vill be accessible for surface and volumetric inspection. The internal surface can be inspected by re=ote visual means. h.k.1.3 Steam Generator - The external surfaces cc the steam generatsr are accessible for surface and volumetric inspection. The reactor coolant side of the steam generator can be inspected internally by remote visual means by removing the manvay covers in the steam generator heads. Portions of the internal surface of the shell can be inspected by removal of M handhole and manway covers. l l l 3 /;8 g , 1 1 h-25 J l l

l D-B h.h.l.h Reactor Coolant Pumps The external surfaces of the pump casings are accessible for inspection. The , internal surface of the pu=p suction is available for inspection by removing the pump internals. 1 4.h.1.5 Pining l The reactor coolant piping, fittings, and attachments to the piping external to the primary shield vill be accessible for external surface and volumetric in-spection. h.h l.6 Dissimilar Metal and Tepresentative Welds All dissimilar metal velds with the exception of Inconel to stainless steel pipe velds will be made in the manufacturer's shops. All dissimilar metal velds will be accessible for inspection during the service life of the nuclear uAit. Dissimilar metal velds on the reactor vessel include only tra core flooding lines, in-core instru=entation guide tubes, and control rod drive housings. f Dissimilar meta. velds in the piping include only attachments and extensions for velding to tae reactor coolant pump suction and discharge nozzles. Dissimilar metal velds in the pressurizer include the surge line, the relief valve and spray line connections. Dissir J =etal velds on the steam generator occur only at the small drain lines and instrument attachments. Representative longitudinal and circumferential velds on the piping, steam gen-erator, pressurizer, and pump casing vill be inspectable as described above. Representative velds on the reactor vessel closure head will be inspectable. Longitudinal and circumferential veld areas on the reactor vessel interior sur-faces vill be inspectable. h.h.l.7 Inspection Schedule The schedule for the type and frequency of inspection in each of the areas mr.n-tiened above vill be established during the detailed design. h.h.2 REACTOR COOLANT SYSTEM TESTS The assembled reactor coolant system vill be tested during final nuclear unit construction and initial startup phases, as follows: h.h.2.1 Reactor Coolant System Precritical and Hot Functional Test This test demonstrates satisfactory preliminary operation of the entire system and its individual components, checks'and evaluates vperating procedures, and determines reactor coolant system integrity at normal ouerating temperature and pressure. h-26 >

D-B , h.h.2.2 Pressurizer Precritical Operational Test This test demonstrates satisfactory preliminary operation of the pressurizer and its individual ecmponents. Spray valve adjustments and heater control adjustments are tested. h.h.2.3 Relief Valves Test In this test all relief valves are set and adjusted, and operating precedures are evaluated. h.k.2.h Unit Power Startup Test This test determines performance characteristics of the entire unit in short periods of operation at steady-state power levels; h.4.2.5 Unit Pcwer Heat Balance This test determines the actual reactor heat balance at various power levels to provide the necessary data for calorimetric calibration of the nuclear instrumentation and reactor coolant system flow rate. h.h.2.6 Unit Power Shutdevn Test This test checks and evaluates the operating procedures used in shutting down the unit and htermines the overall unit operating characteristics during shutdown operations. These tests are in addition to the tests in compliance with code requirements. h.h.3 MATERIAL IRRADIATION SURVEILLANCE Surveillance specimens of the reactor vessel shell section material are installed between the core and inside vall of the vessel shell to monitor the NDTT of the vessel material during operating lifetime. The type of specimens included in the surveillance program vill be Charpy V-notch (Type A) and tensile specimens for measuring the changes in material properties, resulting from irradiation. This is in accordance with ASTM E185 66, "Recc== ended Practice for Surveillance Tests on Structural Materials in Nuclear Reactors." Refer to BAW-10006,

        " Reactor Vessel Material Surveillance Program", for a ecmplete description of the surveillance program.

The reactor vessel material surveillance program vill utilize a total of three 3 surveillance specimen holder tubes located close to the inside reactor vessel vall. The locations of the holder tubes are shown in Figures 3 h5 and 3 h6. In these positions , the specimens will receive approximately 1-3/h times as much radiaition as the reactor vessel receives. The material frem the reactor vessel vill have its initial NDTT determined by the Charpy V-notch impact tests. The predicted shift or change in the NDTT of the vessel material resulting from irradiation is discussed in detail in

 ,      h.l.h l.
   ~

l' l 350 h-27 Arendment No. 3

D-B A

                                                                                                 )

The influence of neutron irradiation on the reactor vessel material properties vill be evaluated by specimen testing as discussed in BAW-10006. The evalu-ation vill be accomplished by testing samples of the nr.terial from the reactor vessel which are contained in the surveillance specimen capsules. These capsules contain steel coupons frcm forgings, veld, and heat-affected zcne material used in fabricating the reactor vessel. Desimeters are placed with the Charpy V-notch ' impact specimens and tensile specimens . The dosimeters vill permit evaluation of flux as seen by the specimens and vessel vall. To-prevent corrosion the specimens are enclosed in stainless steel sheaths. The irradiated semples are tested to deter =ine the material properties , such as tensile or 1'mact, and the irradiated NDTT vhich may be measured in a 3 manner sis 2- 'M initial NDTT. These test results can be compared with the then-ev < on the effects of neutron flux and spectrum on engi- . neering mat . The measured neutron flux and NDTI' may then be compared with the initial NDTT and the predicted NDTT shift to mtnitor the progress or radiation-induced changes in the vessel materials. As the end of reactor design life nears, a significant increase in measured NDTT in excess of the predicted NDTT shift could be investigated by reviewing the vessel stress analysis and operating records . If necessary or required in accordance with the advanced knowledge available at that time, the vessel transient limitations en pressure and temperature may be altered so that vessel stress limits, as stated in h.l.h.3 ) for heatup and cocidown, are not exceeded. h.5 QUALITY CONTROL h.5.1 GENERAL k.5.1.1 Dimensienal Inspection In-process and final dimensional inspections are made to insure that parts of assemblies meet the drawing requirements, and an "as-built" record of these

                                                             .551 gS h-27a                        Amendment Jo. 3

D-B dimensions is kept for reference. A temperature-controlled gauge room is maintained to keep all measuring equipment in proper calibration, and personnel supervising this work are trained in formal programs sponsored by gauge equip-ment manufacturers. 4.5.1.2 Nondestructive Testing The pri=ary purpose of nondestructive testing is to locate, define, and deter-mine the size of material defects to allow an evaluation of defect acceptance, rejection, or repair. Repaired defects are similarly inspected as required by applicable codes.

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

Radiography incorporates the following techniques: . (1) All velds are properly prepared by chipping and grinding valleys between stringer beads so that radiographs can be properly in-terpreted. (2) All radiographs required by code are reviewed by qualified per-sonnel who are trained in their interpretation. (3) An 0.080-in. lead filter is used at the film to absorb " broad-beam scatter" when using high-voltage equipment (above 1 MeV). (h) Fine grain or extra fine grr.in film is used for all exposures. (5) Densities of radiographs are controlled by densitometers. (6) Double film technique is used on all ga==a-ray exposures as well as high-voltage exposures. (7) Films are processed through an automatic processor which has a controlled replenishment, te=perature, and process cycle. (8) Energies are controlled so as to be in the optimum range.

b. Ultrasonic Inspection Ultrasonics is used to examine all pressure-integrity raw =aterial, and pressure-containing welds as required by applicable codes.

Ultrasonic inspection is used as follows: - (1) In addition to being radiographed, pressure containing velds are ultrasonically inspected, where required by applicable codes. ! (2) In order to detect 1sninations which lie parallel to the surface, plates are also inspected by a longitudinal (or normal) beam h-28

                                                            }}2

D-B _ technique. The shear wave is more effective in detecting de-

 )                     fects oriented perpendicular to the surface.

(3) Personnel conducting ultrasonic inspections are trained and qualified.

c. Magnetic Particle Examination Magnetic particle examination is used to de'_act surface or near sur-face defects in machined weld grooves prior to velding, completed veld surfaces, reactor vessel closure studs, and the co=plete ex-ternal surface of vessels including veld seams after final heat treat-ment.

Personnel conducting magnetic particle examinations are trained and qualified.

d. Licuid Penetrant Examination .

Liquid penetrant examination is used primarily to detect surface defects in the veld-deposited cladding and in non-magnetic materials. Personnel conducting liquid penetrant examinations are trained and qualified. h.5.1.3 Weldine and Heat Treating Processes A program of procedure qualification of all velding and heat treating processes which could affect =echanical or metallurgical properties of the material turing fabrication is conducted. Based upon these qualified procedures, production velding and inspection procedures are prepared. All materials are certified upon receipt to be within the same material specification limits as those used to qualify the procedure. The purpose of this program is to establish the prop-erties of the material as received, and to certify that the mechanical proper-ties of the materials in the finished vessels are consistent with those used in tae 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 fabricatiori.

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

c. Subjecting qualifiestion test plates to destructive tests to establish:

Tensile strength. Ducti?.ity. Resistance to brittle fracture of the veld metal, base metal, and heat-affected zone (HAZ) metal. h-29 353

D-B All velders are qualified or requalified, as necessary, in accordnace with the requirements of Babcock & Wi]cox and applicable codes. Each lot of velding electrodes and fluxes is tested and qualified before rePase to insure that / required 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 veld. All velding electrodes and fluxas are maintained dry and free from contamination before use. Records are main-tained and reviewed by velding engineers to insure that approved procedures and materials are being used. Records are maintained for each veld joint and include the velder's name, essential veld parameters, and electrode heat or lot number. h.5.1.4 Material Identification All pressure boundary materials are permanently identified. The identity is maintained throughout =anufacture so that each piece can be located in the finished vessels. h.5.2 REACTOR VESSEL

• The reactor vessel vill be designed, manufactured, and tested in accordance with Section III of the ASME Code and vill be subjected to the following non-dest;uctive tests:

a. All pressure boundary plate vill be ultrasonically inspected by both normal and shear ~ wave.

b. All pressure boundary forgings will be ultrasonically inspected. N

c. )

Carben and lov alloy base material shall be magnetic particle in- ' spected after quench and temper.

d. All accessible velds shall be either magnetic particle or liquid penetrant inspected after hydrostatic test.

e. The external surface o the entire vessel, including veld seams, vill be magnetic particle inspected after final hydrotest.

f. Weld deposit cladding vill be inspected by the liquid penetrant method.

g. All cladding shall be liquid penetrant and ultrasonically tested after stress relief at the highest stress relief te=perature to be employed. ,

h. After fabrication velds vill be inspected by raulography, liquid penetrant, or magnetic particle methods.

i. Stud forgings will be inspected for flaws by two ultrasonic inspec-tions, axial and radial. In addition to the ultrasonic tests, mag-netic particle inspection vill be performed on the finished studs.

T t z 354 i k-30 l

                                                                                .      1

D-B m 4.5.3 i AM GENERATOR

     )

The steam generator will be designed, manufactured, and tested in accordance with Section III of the ASME Code and will be subjected to the following non-destructive testing during manufacture: a, b, d, e, and f - Same as under k.5.2 above. 1

g. All cladding will be liquid penetrant tested after stress relief at the highest stress relief te=perature to be employed.

h. All weld preparations shall be magnetic particle or' liquid penetrant inspected prior to velding,

i. Tube-to-tubesheet velds will be liquid-penec ant examined. These velds will be gas-leak tested prior to final stress relief.

J. Each tube will be inspected ultrasonically and by eddy current. Liquid penetrant examination on the inside and outside of each end of a tube sa=ple from each lot of tubes will be made.

k. After fabrication velds will be inspected radiographically and/or ultrasonically, as applicable. l1 4.5.4 PRESSURIZER The pressuriser vessel vill be designed, manufactured, and tested in accordance with Section III of the ASME Code. It will be subjected to the following non-destructive testing during manufacture:

a,b,c,d,e, and f - Sa=e as under 4.5.2 above.

g. All cladding will be liqi'.d penetrant tested after stress relief at the highest stress relief temperature to be employed.

h. After fabricati,n welds will be inspected radiographically and/or ultrasonically, 0.s applicable.

   ,       h.5 5         REACTOR COOLANT PIPING The reactor coolant piping vill conform to ANSI Code Section P31.7 and to Sec-tion IX of the ASME Code a d, during fabrication, will be subjected to the fol-l3 lowing nondestructive testing:

a , b , e , and f - Same as under 4. 5 2 above .

d. After fabrication velds will be inspected radiographically and I <

by ma6netic particle examinatk ., as required by applicable codes. 1

e. Weld seams and significant weld repairs shall be magnetic-particle or liquid-penetrant inspected after final hydrostatic test.

      .f ,        g. Bond between cladding and base material vill be ultrasonically in-ty              spected on elbows only, h-31             355      Amendment No. 3

D-B

h. All cladding shall be liquid penetrant tested after stress relief at the highest stress relief temperature to be employed.

1. All veld preparations shall be magnetic particle or liquid penetrant inspected pr_ar to velding.

h.5.6 REACTOR COOLANT PUMP CASINGS The pump casings will be manufactured in accordance with Section III of the ASME Code, where applicable, and will be subjected to the following nondestructive tests:

a. The rough casing vill be inspected radiographically,

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

c. The upper and lower stuffing boxes will be inspected radiographically.

d. Welds will be inspected radiographically and by liquid penetrant methods.

k.5.7 VALVES The valves vill be manufactured to ANSI B16.5, inspected in accordance with ANSI B31.7, where applicable, and vill be subjected to the following nonde-3 structive tests: ,

a. Valve body rough castings will be inspected radiographically.

b. These finished castings will be liquid penetrant examined.

c. Valve body forgings vill be inspected by ultrasonics and liquid penetrant examinations. h.6 REFERENCES

1. Hjarne, L. , and Leimdorfer, M. , "A Method for Predicting the Penetration and Slowing Down of Neutrons in Reactor Shields," Nuclear Science and Engineering 2h, pp 165-lTh, 1966.

2. Cadwell, et al., The PDQ-5 and PDQ-6 Programs for the Solution of the Two-Dimensional Neutron Diffusion-Depletion Problem, WAPD-TM h77, January 1965

3. Aalto, et al. , " Measured and Icedicted Variations in Fast Neutron Spectrum in Massive Shields of Water and Concrete," Nuclear Structural Engineering 2, pp 233-2h2, Au6ist 1965

h. Avery, A. F., The Prediction of Neutron Attenuation in Iron-Water Shields, AEEW-R 125, A ril 1962.

5 Clark, R. H., and Baldwin, M. N., Physics Verification Program, Part II, BAW-36h7 h, June 1967 Amendment No. 3 h-32

D-B

 ~

6. Forse, L., Reactor Vessel Design Considering Radiation Effects, ASME Paper No. 63-WA-10,0.

T. Pellini, W. S. and Puzak, P. P., Fracture Analysis Diagram Procedures for the Fracture-Safe Engineering Design of Steel Structures, Welding Research Council Bulletin 88, May 1963.

8. Robertson, T. S. , Propagation of Brittle Fracture in Steel, Journal of Iron and Steel Institute, Volume 175, December 1953.

9 I;ihara, H. and Masubichi, K., Effects of Residual Stress on Brittle Fracture, Welding Journal, Volume 38, April 1959

10. Miller, E. C., The Integrity of Reactor Pressure Vessels, ORNL-NSIC-15, May 1966.

Babcock & Wilcox Tonical Reports BAW-10002 Research and Development Report fer the Once-Through Steam Generator BAW-10006 Reactor Vessel Material Surveillance Program BAW-10018 Analysis of the Structural Integrity of a Reactor Vessel Subjected to Thermal Shock 8 . W v ~ 357 h-33

D-B Table h-1 Tabulation of Reactor Coolant System Pressure Settings Item Pressure, usig Design Pressure 2,500 Operating Pressure 2,185 Code Relief Valves 2,500 Pilot-Act. Relief Valve 2,300 High-Pressure Trip 2,350 High-Pressure Alarm 2,285 Low-Pressure Alcrm 2,085 Low-Low-Pressure Alarm 1,850 . Low-Pressure Trip 1,750

                                                             .. /

I a 358 h-3h

D-B Table h-2 Reactor Coclant Quality Parameter Value Total Solids, Max. (Including Dissolved 1.0 and Undissolved But Excluding H 3303 ""d LiOH), ppm Baron, ppm See Figure 3-1 I Lithium as Li, ppa 0.5 - 2.0 (Equivalent range as ILiOH is 1.h55 to 5.82, ppm) pH at 77 F h.8 - 8.5 (Equivalent pH at 600 F is 6.8 to T.8) , 02 (Max.), ppm Not Applicable (With Proper H2 Specification at Critical Condi-tion, Dissolved 02 is Assumed Not to be Present) C1 (Max.), ppm 0.1 H , Std cc/l 2 15 - h0 F (Max.), ppm 0.1 Hydrazine (Required During Shutdown), ppm 8 - 25 [6

iS9 h-35

I D-B Table h-3 Reactor Vessel Design Data O) I+em _ Data Design / Operating Pressure, psig 2,500/2,185 Hydrotest Pressure (Cold), psig 3,125 Desi'gn/ Operating Temperature, F 650/600 Overall Heig..t of Vessel and Closure 39-0 Head, ft-in. Straight Shell Thickness, in. 8-T/16 Water Volume, ft3 h,o5g Thickness of Insulation, in. 3 Number of Reactor Closure Head Studs 60 . Flange ID, in. 165 Shell ID, in. 171 Inlet Nozzle ID, in. 28 Outlet Nozzle ID, in. 36 Core Flooding Water Nozzle ID, in. 11-1/2 m

                                                                \
                                                               )

Table h-h Pressurizer Design Data Item Data Design / Operating Pressure, psig 2,500/2,185 Hydrotest Pressure (Cold), psig 3,125 Design / Operating Temperature, F 670/650 Normal Water Volume, ft3 800

• Normal Steam Volume, ft3 Too Surge Line Nozzle Diameter, in. 10 Overall Height, ft-in. hh-11-3/h s

m o h-36

D-B Table h-5 Steam Generator Design Data Item Data per Unit Design Pressure (Reactor Coolant / Steam), psig 2,500/1,050 Hydrotest Pressure (Tube Side-Cold, Reactor 3,125 Coolant), psig Design Temperature (Reactor Coolant / Steam), F 650/600 Reactor Coolant Flov, lb/hr C3.6 x 106 Heat Transferred, Btu /hr h.52 x 109 Steam Conditions at Rated Load, Outlet Nozzles: Steam Flow, lb/hr 5.68 x 106 Steam Temperature, F 570 (2 Superheat) Steam Pressure, psig 910 Feedvater Temperature, F 455 Overall Height, ft-in. 73-2-1/2 Shell OD, in. 151-1/8 Reactor Coolant Water Volume, ft3 2,030 l l s . w 34/ . p k-37

D-B Table 4-6 Steam Generator Feedvater Quality Parameter Value pH at 77 F 9.3 - 9.5 (NH3 ) Dissolved 02 (Max.), ppm 0.007 SiO2 (Max.), ppm 0.02 Fe (Max.), ppm 0.01 Total Solids (Max.), ppm 0.05 Cu (Max.), ppm 0.002 l

                                                       .l l

1 t

                                            . - g h-38 J

W e

1. +++% %'+4 , eEEv <e 1,em TEST TARGET (MT-3) 1.0 5 Lu E!4

                         ..                 5 m gn m    m l.1             = HM 11 1.25          l.4    i  1.6 I

c 6" d i MICROCOPY RESOLUTION TEST CHART 1 4llI//% ++

 ?I 'N                                                               h.

y,,,/ 4)p,g*. 1 i

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

4 4 '*kh ' TEST TARGET (MT-3) k.h 10 " HE m EosE=E m i l,l [" lllM

                       -                       1.8 1.25         1.4       1.6 1

4 6" MICROCOPY RESOLUTION TEST CHART O \

• %[f65,,I .

                                                                 'itAf N-.     ._..u-       - - - _ = ,- _ - . _ . b :=2a .

D-B Table h-7 Reactor Coolant Pump Design Data Item - Data per Unit Number of Pumps k Design Pressure, psig 2,500 Hydrotest Pressure (Cold), psig 3,3h0 Design Temperature, F 650 Operating Speed (Nominal), rpm 1,180 Pumped Fluid Temperature, F 60 to 580 Developed Head, ft 355 2 Capacity, spm 88,000 Hydraulic Efficiency, % 86 Seal Water Injection (Max.), gpm 15 per pump 2 Seal Water Return (Max.), gpm 15 per pump Pump Nozzle ID, in. 28 Overall Unit Height, ft 26 2 Water Volume, ft3 loo Motor Stator Frame Diameter, ft 8 Pump-Motor Moment of Inertia, lb-ft2 70,000 Motor Data: Type Squirrel-Cage Induction, Single Speed Voltage 13,200 Phase 3 l2 Frequency, cps 60 Starting Across-the-Line Input (Hot Reactor Coolant), kW 5,600 InI.ut (Cold Reactor Coolant), kW 7,h00 v e - h h-39 Amendment No. 2

D-B Table k-8 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 lho Design / Operating Pressure, psig 2,500/2,185 Hydr est Pressure (Cold), psig 3.125 Design / Operating Temperature, F 650/608 Design / Operating Temperature 670/650 - (Pressurizer Surge Line), F Water Volume, ft3 1,91o  : i 5 k-k0

D-B

                                            Table h-9
       ,                                  Transient Cycles 3

(h0 Year Design Life) Design Design Estimated Description Frequency Cycles Actual Cycles

1. Heatup, 70 to 557 F; 0.5/ Month 2h0 80 . 2 Cooldown, 557 to 70 F

2. Heatup, 532 to 582 F; 3/ Month lhho 770 3 Cooldown, 582 to 532 F (i 3 Ramp Loading and Unloading 3/ Day h8,000 36,000 (15 - 100 - 15%)

h. Step Loading Increase (10%) h/ Week 8,000 6,000

5. Step Leading Decrease (10%) h/ Week 8,000 6,000

6. Step Load Reduction to h/ Year 160 120 Auxiliary Load, 100 to 5%

7 Reactor Trip From Rated 10/ Year 400 300 Power

8. Rapid Depressurization 2 ' fear 80 LO 9 Sudden Change in Coolant 1/2/ Year 20 10 Flow

( 10. Rod Withdrawal Accident 1/ Year h0 1

11. Loss of Coclant 1/h0 Year 1/2 0 High Pressure Injection 1/h0 Year 1/2 0 Low Pressure Injection 1/h0 Year 1/2 0

12. Miscellaneous Transients --

20 10 (Including Hydrotests) 1: 003

,-4 h kl                     Amendment No. 6

D-B Table !+-10 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 ASME " III as Applicable Class A Casing 3 Motor IEEE(b) , NEMA (* } , and ANSI (d) Piping e.nd Valves ANSI (*) B31.7 - 1968 as 3 f Applicable , (a)A=erican Society of Mechanical Engineers, Boiler and Pressure Vessel Code. Section III covers Nuclear Vessels. Institute of Electrical and Electronics Engineers. (C National Electrical Manufacturers Association. (d) American National Standards Institute No. C50.2 - 1955 and '] 3 C50.20 - 195h. s American National Standards No. B31.7 dated February 1968, including Errata sheet dated June 1968. 4 Amendment No. 3 h h2 [

D-B

 . <-                                                Table h-11 Materials of Constructicn Component                   Section                          Material
                -Reactor Vessel      Pressure Plate                   SA-533, Grade B, Class II ")

Pressure Forgings A-508-6h, Classes 1 and 2 (Code Case 1332-h) Cladding 18-8 Stainless Steel Thermal Shield and Internals SA-2h0, Type 30k, and Inconel-X Nozzles A-508-6h, Classes 1 and 2 (Code Case 1332-h) ~ Inconel-SB167 Control Rod Drive Mechanism Inconel-SB167 and In-Core Instrument Penetrations Steam Generator Pressure Plate SA-516, Grade TO SA-533, Grade B, Class 1 Pressure Forgings A-508-6h, Class 1 (Code Case 1332 h) Cladding for Heads 18-8 Stainless Steel Cladding for Tubesheets Ni-Cr-Fe Tubes SB-163 Nozzles A-508-64, Class 1 (Code Case 1332 h); SA-106, Grade B; SA-312, Type 30h Pressurizer Shell, Heads, and External SA-516, Grade To Plate Pressure Forgings A-508-6h, class 1 , (Code Case 1332-h) Cladding 18-8 Stainless Steel .

       . c, . ,                                                                 005 6

(a)This material is metallurgically identical to SA-302, Grade B, as modified

   ~~

by Code Case 1339 h h3

D-B Table h-11 (Cont'd) Component Section Material Internal Plate SA-2h0, Type 30h Internal Piping SA-312, Type 30h Nozzles A-508-6h, Class 1 (Code Case 1332-k); SA-106, Grade B; SA-312, Type 30h Reactor Coolant 28 in. and 36 in. SA-516, Grade 70 (Elbows) Piping A-106, Grade C (Straights) Cladding 18-8 Stainless Steel 10 in. A-h03, Grade WP-316 (Elbows) A-376, Type 316 (Straights) Nozzles A-508-6h, Class 1 (Code case 1332-h); A-336, Class F8M A-182, Type F-316 Y p l k-kh

D-B Table h-12 References for Figure h-b - Increase in Transition Temperature Due to Irradiation Effects for A302B Steel Neutron Ref Temp, osure, NDTT, No. ' Reference Material Type F n/cm (>l MeV) F 1 ASME Paper All Steels Max. Curve for 550 Data No. 63-WA-100 (Figure 1). 2 ASTM-STP 380, A302B Plate Trend Curve for 550 Data p 295 3 NRL Report 6160, A302B Plate 550 5 x 1018 65 p 12. h ASTM-STP 3hl, A302B Plate 550 8 x 1018 85I *} p 226. 5 ASTM-STP 3hl, A302B Plate 550 8 x 1018 100 p 226. 6 ASTM-STP 3hl, A3023 Plate 550 1.5 x 10 19 130(*) p 226. ? 7 ASTM-STP 3hl, A302B Plate 550 1.5 x 1019 1h0 p 226. 8 Quarterly Report A302B Plate 550 3 x 1019 120 of Progress,

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

1-31-65 9 Quarterly Report A302B Plate 550 3 x 1019 135 on Progress,

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

1-31-65 l N 007 1 (a) Transverse specimens. l h h5

D-B Table h-12 (Cont'd) Heutron Ref Temp, Exposure, NDTT, Wo. Reference MateriaJ. Type F n/cm2 ( 1 MeV) F 10 Quarterly Report A302B Plate 550 3 x 1019 1ho of Progress,

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

1-31-65. 11 Quarterly Report A3023 Plate 550 3 x 1019 170 of Progress,

       " Irradiation Ef-fects on Reactor                                                            ~

Structural Mate-rials," 11-1-64/ 1-31-65 12 Quarterly Report A302B Plate 550 3 x 1019 205 of Progress,

       " Irradiation Ef-fects on. Reactor Structural Mate-                                                        N i

rials," 11-1-6h/ M 1-31-65 13 Welding .-asearch A302B Weld 500 5 x 1018 70 Supplemer. c., Vol to 27, No. 10, Oct 575 1962, p 465-S. 14 Welding Research A302B Weld 500 5 x 1018 So Supplement, Vol to 27, No. 10, Oct 575 1962, p h65-S. 15 Welding Research A302B Weld 500 5 x 1018 37 Supplement, Vol to 27, No. 10, Oct 575 1962, p 465-S. 16 Welding Research A302B Weld 500 5 x 1018 Supplement, Vol 25 to ' 27, No. 10, Oct 575 1962, p h65-S. j ~-.g k h6

D-B Table h-13 Stresses Due to a Maximum Design Steam Generator Tube Sheet Pressure Differential of 2,500 usig at 650 F Stress Computed Value Allowable Value Primary Membrane 22,000 psi 37,200 psi (0 9 Sy) Primary Membrane Plus 39,700 psi 55,900 psi Primary Bending (1.35 Sy ) Table h-lh Ratio of Allowable Stresses to Co=puted Stresses for a Steam , Generator Tube Sheet Pressure Differential of 2,500 psig Component Part Stress Ratio Primary Head 4.02 Primary Head Tube h.02 Sheet Joint Tubes 1.07 Tube Sheet Max Avg Ligament 1.02 Effective Ligament 1.70 009 \ h-47

r T. _

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e FIGURE 4-1 AMEN 0 MENT NO. 7

PRES 3URIZER

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i 1 i 41'-0* . 34'-0" i i t t, wm DAVIS BESSE NUCLEAR POWER STATION REACTOR COOLANT SYSTEM ARRANGEMENT - ELEVATION 031,3 FIGURE 4-2 AMENDMENT NO. 2 ,

                              .7 ----- -      3.   ~ - _, _ _ ,_ _, . _                             , , ,                                   , , _

t e 21'-0"  ; 13'-if"- PRESSURIZER

                                    +

( s 7'-6g c- s l D7 N 30 N

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                   ,                             s                                          1 32'-0"
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                                %*J DAVIS-BESSE NUCLEAR POWER STATION REACTOR COOLANT SYSTEM ARRANGEMENT ' PLAN        3 OO j,}~                        FIGURE 4-3 AMENDMENT NO. 2

l l( t 1 4 6 1 2 4 6 1 2 4 6 500 - ' ( .;

                                          ~

0 400 / e Maximum Curve For [

                                          =             550 F Data
                                         $              (See Reference 1,3 2   300 E
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e

                                                                                                                                    /        Trend Cu ve g    200                                                12 '                    ,/

5 0 F Data a

                                        =                                                       11. /                                        ( S e e R e f e r e n c e 2.)
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C 6 O 5. /

                                                                                            /s9 100 Z                              11       +4
                                                                                    /

14+ E[ 16++ 0 i i ii iiiiinii i i iiii,e iii i i i....iii 10 18 10 19 1020 1021 l Integrated Neu t r on E x posu r e (E> Me v), n/cm2 Notes 1 All data is for 30 ft-lb "Fix"

2. Numbers on Curves lndicate References in Table 4-11 DAVIS-BESSE NUCLEAR POWER STATION Nil-DUCTILITY TRANSITION TEMPERATURE INCREASE VERSUS INTEGRATED NEUTRON EXPOSURE FOR A3028 STEEL FIGURE 4-4 GD

n. - _ - - _ - __ __ _

20'- 9" (ACROSS N0ZZLE ' ACES) pwem f M 6 l 3 i M w e ,kti 3

                                )                                                   .

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                            /
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Mb l dd ' ne ,

                                          %(

w DAVIS BESSE NUCLEAR POWER STATION REACTOR VESSEL FIGURE 4-5

SAFETY VALVE N0ZZLE ~ VENT N0ZZLE l (TYPICAL 0F3) t

 .          SPWAY N0ZZLE        o
                                     /                                                                                                                              i SPRAY LINE MOZZLE                                              l
                                         /\

1 l LEVEL SENSING N0ZZLE

                               \                             \

(TYPICALOF3) N \

                               \                             \

N \

                               \                             \
                               \                             \

7- STEAM SPACE N > NORMAL WATER \ LEVEL \

                               %                            N N                            N
                               \                            \

VESSELSUPPORTS(8)

                               \                            \

x \ N \

                              \                             \
                              \                             \

THERM 0WELL '

                              \                                                                     SAMPLING N0ZZLE y-                           ch   [
                              \                             \                                          HEATER BUNDLE
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LEVEL SENSING MOZZLE y -

                                                 /            /

(TYPlCAL OF 3) l T' _

                                         =

SURGE LINE N0ZZLE DAVIS-BESSE NUCLEAR POWER STAT

                                  .              O{.tj {3                                                      PRESSURIZER FIGURE 4 6

REACTOR COOLANT INLET G

                                         /y  /            d                                  - MA N WAY J

V,/ s bkkk kk Ni ;i 1 p >:jj:,1 a , .!;st c ,

                          'I f ) f. lf) -['lI !;' r,  .
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                           $pp$l,:hid              '*'        --

iji k D] , d!f'U!p:,, f i' " o I i ,i. c i j[ .

                                                                                  ,h*"                           -

e y 74 c. , 'p q y!' .

• f - FEEDwATER INLET 4I'-tM.. ' ..v.;. .: I40 t Mh'0l;d.ij ! l d '

                     @(,@de.:v!,'i;u'!;ll'fp se ., ,,                 ,=13, l t,      f                  !.

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OUTLET - O R AIN INSPECTION PORT STEAM GENERATOR DAVIS-BESSE NUCLEAR POWER STATION ORi q, STEAM GENERATOR FIGURE 4-7

I N L ET F l l O O SUPERHEAT REGION AT RATED LOAD q q ' a a 8AFFLE p # FILM B0ILING p g REGION AT + k - STEAM OUTLET RATED LOAD \ # g , BYPASS STEAM FOR g, g FEEDWATER HEATING FEEDWATER INLET DOWMCOMER j SECTION NUCLEATE BOILING REGION SHELL l AT RATED LOAD u o [

                                                                     .                          1 o                 U                 U REACTOR COOLANT GUTLET l              DAVIS-BESSE NUCl. EAR POWER STATION 00J8 STEAli GENERATOR HEATING REGIONS FIGURE 4-8

100 100 g Downcomer j 90 Superheat 90 Level . 80 ' ,# 80 l h

                    ,     70                                                                                                               70 e

J 60

                                                                                          /
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                      $                                                             l s 50 v,
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                      %   40

• 40

( 3 30

                                                     /

[ \ 30 $ Mucieate E 20 ' 9 20 Film f 10 summ e _" lo O

                                                    #V                          '

O O 10 20 30 40 50 60 70 80 90 100 Rated Power.f. DAVIS-BESSE NUCLEAR POWER STATION STEAM GENERATOR HEATING SURFACE AND DOWNCOMER LEVEL VERSUS POWER FIGURE 4-9 0039

620 600

                                                                                                              -         /

Reactor Coolant y Avg. Tube w 580 - Wall Tup. J c Mean Tube

5. -- -- ---

y-- -M Temp.

                                 =  560                                  >                                                                                  /                          Steam

[ --

                                                                                                                                            /                                          Outlet

( --. _ _ . 540 f ' l " dl - _ _ , / Temp.

                                        /                                          Shell temp.

l 0 20 40 60 80 100 Bot tom Top Tubeshee t Tubesheet Tube Length, 5 of Total DAVIS-BESSE NUCl. EAR POWER STATION STEAM GENERATOR TEMPERATURES FIGURE 4-10 00,20 l

                                                                                         \
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l l 0

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                            \                     SUCTION
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                              \            ,

N0ZZLE DAVIS-BESSE NUCLEAR POWER STATION REACTOR COOLANT PUMP FIGURE 4-11 1 AMENDMENT NO. 2 0021

O 280 MAX ll;UM NOTT SHIFT I '

                                                                                                      /

AT DESIGN nvt j / 240 -

                                                            /
                                                                /,/                40                          -

200

                                                  /                                      i
    '                                           0 160
                                      /         e J                              7            g I                               YEARS OF EQUIVALENT EXPOSURE TIME
    =                               AT 80% LOAD AND 2772 MWt
                             /      (REACTOR PRESSURE VESSEL MATERIAL) 2 i:o                 / 10
                      /                                                                                                    ,

80 f

                /
         .    /

DESIGN VALUE FOR nvt 0 0 0.4 0.8 1. 2 1.6 2.0 2. 4 2.8 3. 2 3. 4 l Integrated Neutron Exposure (E > I Mev) n/cm2 x 3o-19

                                                                            .                         JIFJ a

DAVIS-BESSE NUCLEAR POWER STATION PREDICTED NOTT SHIFT VERSUS REACTOR VESSEL IRRADIATION Figure 4-12 0022 __}}