Text

Chapter 1 of Davis-Besse PSAR, Introduction & Summary. Includes Revisions 1-8
	ML19319C280
Person / Time
Site:	Davis Besse
Issue date:	08/01/1969
From:	; TOLEDO EDISON CO.
To:	;
References
	NUDOCS 8002110780
	Download: ML19319C280 (71)
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i D-B TABLE OF CCNTENTS lection Page 1 INTRODUCTION AND

SUMMARY

1-1

1.1 INTRODUCTION

1-1 1.2 DESIGN HIGHLIGHTS 1-2 1.2.1 SITE CHARACTERISTICS 1-2 1.2.2 POWER LEVEL 1-2 1.2.3 PEAK SPECIFIC POWER LEVEL 1-2 1.2.h CONTAINMENT SYSTD4S 1-3 l.2 5 ENGINEERED SAFETY FEATURES 1.h 1.2.6 ELECTRICAL SYSTEMS AND EMERGENCY POWER 1-6 1.2 7 ONCE-THROUGH STEAM GENERATORS 1-6

,_ 1.3 TABULAR CHARACTERISTICS 1-7 1.4 PRINCIPAL DESIGN CRITERIA 1-17 1.5 RESEARCH AND DEVELOPMENT REQUIREMENTS 1 h6 1 5.1 XENON OSCILLATIONS 1-46 1 5.2 THERMAL AND HYDRAULIC PROGRAMS 1-h7 1 5.3 FUEL ROD CLAD FAILURE 1-h7 1 5.h CONTROL ROD DRIVE LINE TEST l-h9 1.5.5 ONCE-THROUGH STEAM GENERATOR TEST 1-50 -

1 5.6 SELF-POWERED DETECTOR TESTS 1-51 1 5.7 BLOWDOWN FORCES ON INTERNALS AND CORE 1-52 1.6 PROPOSED STATION DESIGN IN THE AREAS OF CONCERN IDENTIFIED IN ACRS LETTERS AS ASTERISKED ITEMS 1-53 1.7 THE TOLEDO EDISON COMPANY COMPETENCE TO BUILD AND OPERATE DAVIS-BESSE NUCLEAR POWER STATION 1-56 1.8 IDENTIFICATION OF CONTRACTORS 1-56 19 CONCLUSIONS 1-56 10 1-1 L. J

D-B LIST OF TABLES 1'

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Table No. Title Page, 1-1 Engineered Safety Features 1-5 l-2 ' Design Parameters - Davis-Besse Nuclear Power Station 1-11 D

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t D-B N-i LIST OF FIGURES

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Figure No. Title 1-1 Site Location Plan-1-2 Site Plan 1-3 Site and Station Profiles 1h Auxiliary Building Roof 1-5 Operating Floor 1-6 Refueling Floor 1-T Ground Floor

. 1-8 Below Grade Floor 1 i

1-9 Basement Floor

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I 1-10 Sections A-A and B-B

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1-11 Sections C-C and D-D I

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D-B 1 INTBODUCTION AND SGIMARY

1.1 INTRODUCTION

This Preliminary Safety Analysis Report is submitted in support of the appli-cation by The Toledo Edison Company and The Cleveland Electric Illuminating Company for a utilization facility construction permit and operating license for the Davis-Besse Nuclear Power Station, Unit No.1 on a site on Lake Erie in Ottava County, Ohio, approximately six miles northeast of Oak Harbor.

Toledo Edison will share undivided ownership as tenants in common, of the Davis-Besse Station with The Cleveland Electric Illuminating Company, with Toledo Edison being responsible for the design, construction and operation of the station and for the prosecution of the license application.

The station will have a pressurized water nuclear steam system furnished by The Babcock & Wilcox Company which is similar in design concept to other projects recently licensed or currently under review by the Atomic Energy Division. The balance of the station will be designed and constructed by Toledo Edison with the assistance of Bechtel Corporation and its affiliate, Bechtel Company, and will be similar in design concept to a number of current projects.

The reactor is designed for a power output of 2,633 E t. The license application rating is the warranted rating of 2,633 E t which will give a net station electrical output of approximately 872 We. All safety systems including containment and engineered safety features are designed and evalua-ted for the expected ultimate output of 2,772 MWt and the analysis of all postulated accidents uses this ultimate power rating. The turbine generator and all stess and power conversion systems are designed for the ultimate power rating and at this rating, the net station output will be approxi-mately 906 MWe. The project schedule is based on fuel loading in June 1974 and commercial operation in December 1974. To meet this schedule, site preparation activities will begin in June 1970, and a construction pemit will be needed in October 1970.

The remainder of Section 1 of this report su-mizes the principal design features, site characteristics and safety features of the Davis-Besse Station. Appendix 1A reviews the technical competence of Toledo Edison, Bechtel and Babcock & Wilcox; Appendix 1B reviews the Quality Assurance program; and Appendix 1C reviews the Project Staff.

Section 2 contains a description and evaluation of the Davis-Besse site and environs, and supports tbe suitability of the site for a reactor of the size and type described. Se. ' ions 3 and 4 describe the reactor and the reactor coolant system. Section 5 describes the containment structure and related systems. Sections 6 through 11 describe the other auxiliary systems.

Sections 5, 6, T, 8, and 9 include descriptions of the various systems m directly related to engineered safety features. Section 12 provides pre-ldminary in*ornation relating to station organization and personnel train-ing. Section 13 d:"cribes, in preliminary form, the approach to the subject of initial te sts and operation by the Applicant. Section 14 relates to safety analysis and sunmarizes the analyses which demonstrate the adequacy of the reactor protection system, the containment and the engineered safety features. Section 14 demonstrates that the consequences of various postu-2 13 1-1

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lated accidents are WP'-*n the guidelines suggested in Federal Regulation 10 CFR 100. Section -rovides identification and justification for selecting the varia' .2 conditions that were detemined from the prelim-inary safety analysis and evaluation to be probable subjects of technical specifications for the station.

This report presents descriptive material and analyses of a preliminary design. As the design work progresses to a final detailed design, the sta-tion description, preliminary design numbers, and the related analyses vill be subject to change and refinement.

1.2 DESIGN HIGHLIGHTS 1.2.1 SIE CHARACERISTICS The Davis-Besse station site of approximately 900 acres is located along the south shore of Lake Erie in Ottava County, Ohio, approximately six miles northeast of Oak Harbor and 21 miles east of Toledo as shown on Figure 1-1.

The eastern half of the site is marshland and vestern portion is farmland.

The site provides a minimum exclusion radius of approximately 2,400 feet.

The low population zone radius is 2 miles and the population center distance is 20 miles. The area immediately around the plant will be built up and protected approximately 16 feet above the mean low water level of Lake Erie.

The terrain around the site is relatively flat and, as a result, the site is well ventilated.

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1.2.2 POWER LEVEL Initially licensed power for the reactor core. is proposed at 2633 Wt Core performance analyses in this report are based on this initial power level.

Operating confirmation of reactor core parameters is expected to support an ultimate core power level of 2T72 wt, and Davis-Besse vill be designed to operate at this output. Postulated accidents that could release fission

,- products to the environment have been analyzed on the basis of 2772 MWt.

An additional 17 MWt will be contributed to the cycle by the reactor coolant -

. pumps, resulting in a net electrical output of about 872 MW at initially licensed power and 906 MW ultimately.

1.2.3 PEAK SPECIFIC POWER LEVEL The peak specific power level in the fuel for initial operation at 2633 MWt results in a maximum thermal output of 17 8 kW per foot of fuel rod. This value is comparable with other reactors of this size and therefore repre-sents no extrapolation of technology. This comparison may be seen in the information presented in Table 1-2.

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D-B g 1.2.k CONTAINMENT SYSTEMS The containment for the station consists of two independent structures supported by a common concrete foundation; a steel containment vessel enclosed within a reinforced concrete shield building, as shown in Figure 5-1.

The containment vessel, including all its penetrations, is a lov leakage steel ar. ell which is designed to withstand a postulated loss-of-coolant accident and to confine a postulated release of radioactive material. Safety features directly associated with the containment vessel are the safety injection system, the containment spray system, the containment cooling system, and the containment isolation system. The containment vessel vill be designed, fab-ricated, erected, and tested in accordance with the ASME Boiler and Pressure Code,Section III, Class B.

The shield building is a concrete structure surrounding the containment vessel and is designed to provide biological shielding from hypothetical accident conditions, biological shielding during normal operation, environmental pro-tection for the containment vessel for adverse atmospheric conditions and ex-ternal missiles, and a means for collection and filtration of fission product leakage from the containment vessel following a hypothetical accident. The shield building ventilation system is an engineered safety feature designed to fulfill this last objective.

The containment vessel is a cylindrical steel pressure vessel with hemispher-ical dome and ellipsoidal bottom which houses the reactor vessel, reactor coolant piping, pressurizer, pressurizer quench tank and coolers, reactor coolant pumps, steam generators, core flooding tanks, let-down coolers related piping systems, and normal and emergency ventilation system. It will be completely enclosed by a reinforced concrete shield building having a cylindrical shape with a shallow dome roof. An annular space is provided between the vall of the containment vessel and the shield building. Clearance is also provided between the containment vessel and the dome of the shield building.

The containment vessel and shield building vill be supported by a concrete foundation founded on firm rock structure. The containment vessel vill be constructed in a two stage operation and in a manner that will conform to the ASME Boiler and Pressure Vessel Code, Article 14, N-lh11.

With the exception of the concrete under the containment vessel there vill be

, no structural ties between the containment vessel and the shield building above the foundation slab. Above this there vill be virtually unlimited freedom for differential movement between the containment vessel and the shield building.

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D-B 1.2 5 ENGINEERED SAFETY FEATURES ,

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Engineered safety features are employed to reduce the potential radiation dose to the general public from the Maximum Hypothetical Accident (lk.2.2.h) below the guideline values of 10 CFR 100. The potential dose is reduced by immediate, automatic isolation of all Containment Vessel fluid penetrations that are not required for limiting the consequences of the accident, there-by eliminating potential leakage paths. Long-term potential releases fol-loving the accident are minimized by rapidly reducing the Containment Vessel pressure to near-atmospheric within 2h hours, thereby reducing the driving potential for fission product escape.

In addition, the engineered safety features will prevent core meltdown should the worst postulated loss-of-coolant accident occur. This is ac-complished by large capacity injection core-flooding systems, parts of which are continuously operated for nomal purposes and are immediately available for emergency duty. These systems, coupled with the thermal, hydraulic, and b? swdown characteristics of this reactor, will reliably prevent excessive metalvater reactions (or core disfiguration into a geometry 7 l vhich could prevent further cooling and allow core melting).

Equipment for the Engineered Safety Features of the nuclear unit, and with its normal operation modes, are as follows:

a. High pressure injection - normally shutdown.

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Core flooding system - Self-operating when emergency conditions )

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require its use. No external signal or power source required for operation.

8l c. Lov pressure injection - normally operates for shutdown cooling as part of the decay heat removal system.

d. Containment Vessel spray system - normally shut down.

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e. Containment Vessel cooling system - normally operating, except for emergency cooling vatar supply.

f. Containment Vessel isolation system - operates on test or accident signal.

Table 1-1 lists equipment supplied for the engineered safety features.

1-k Amendment No. 8 16

. ..~.~ = ..- m D-B Table 1-1 Engineered Safety Features Function Total Equipment Installed High Pressure Injection 2 pumps 1 storage tank l6l8 Core Flooding System 2 tanks l8 Low Pressure Injection 2 pumps (decay heat removal) 2 heat exchangers l8 Containment Vessel Spray System 2 pumps and 2 spray headers Containment Vessel Cooling System 3 air recirculating and cooling units

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D-B 1.2.6 ELECTRICAL SYSTEMS AND EMERGENCY POWER The Davis-Besse Station, Unit No.1, vill be a part of The Toledo Edison system and a pool unit for the CAPCO Group. The CAPCO members will be inter-connected by 3h5 kV lines and Toledo Edison is interconnected with the Michigan Pool to the north and the American Electric Power Company system to the south by 345 kV lines.

The Davis-Besse Station vill have three 345 kV lines into the station to provide redundant sources of outside power and to provide means of trans-mitting electrical output into the Toledo Edison system and the systems of .

the CAPCO Group.

There vill be two redundant startup transformers fed from separate 3h5 kV bus sections. Each startup transformer vill have adequate capacity to 2 l supply the complete auxiliary system but will normally feed one 13.8 kV bus and provide backup for the other one. Additional auxiliary power voltage levels vi31 be 4160 V, 277/h80 V, 120/208 V and 125/250 V d.c.

Two redundant emersency diesel generators will supply emergency power for the engineered safety features and other components as required.

These normal, standby, and emergency sources of auxiliary electrical power assure a safe and orderly shutdown of the unit. They also assure the ability to maintain a safe shutdown condition under all credible cir-cumstances.

1.2.7 ONCE-THROUGH STEAM GENERATORS The design of the steam generators is based on ext.ensive research, develop-ment, and experimental work on boiling heat transfer performed by B&W over the past 11 years. Each generator is a vertical shell-and-tube, counter-flow heat exchanger with reactor coolant on the tube side and steam on the shell side. Feedvater is pumped into the generator, heated to saturation by direct mixing with steam, converted to stc m and superheated in a single pass through the generator. The basic design parameters, such as feedvater heating, boiling length, superheat length, and performance characteristics, have been confirmed by testing a full-length, 7-tube unit and a 37-tube unit. Tests are continuing to provide additional data in these design areas for the 37-tube test unit. .In addition, testing will continue with one 19-tube full-length unit and two parallel 19-tube units.

With the once-through design, natural-circulation flow is adequate to re-move full decay heat without the use of reactor coolant pumps. Even with total loss of pumps, no fuel rod vill experience departure from nucleate boiling.

18 Amendment No. 2 1-6

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1.3 TABULAR CHARACTERISTICS Table 1-2 is a comparative list of important design and operatlag character-istics of Davis-Besse Unit 1, Midland Units 1 and 2, (Consumers Power Com-pany), Oconee Units 1, 2, and 3 (Duke Power Company), and Beaver Valley Unit.

The design and operating parameters of the Midlend, Oconee, and Beaver Valley units are close to those of Davis-Besse. Davis-Besse has a slightly higher rated core power than the Midland & Oconee units, but is a near duplicate in other respects. The data in Table 1-2 represents information presented in available station descriptions and in Safety Analysis Reports submitted for licensing.

The design of each of these stations is based on information developed from operation of commercial and prototype pressurized water reactors over a number of years. The Davis-Besse unit design is based on this existing power reactor technology and has not been extended beyond the boundaries of known .

information or operating experience. '

The similarities and differences of the features of the reactor units listed in Table 1-2 are discussed in the following paragraphs. In each case, the item number used refers to the item numbers used in the table.

Item 1. Hydraulic and Thermal Design Parameters The rated power of Davis-Besse is slightly higher than that for the Oconee and tidland units. The slight variation in other parameters between Davis-Besse and the other B&W units is due to the slight difference in power level.

The reactor coolant flow rate and operating pressure are the same for C e Midland, Oconee, and Davis-Besse units. The conservatism of design of Davis-Besse is evidenced by the DNBR of 1.50 (W-3, at the overpower condition.

Item 2. Core Mechanical Design Parameters The table presents comparable mechanical design data for the canless fuel assembly for the B&W units, and the canless fuel assembly used for the Beaver Valley unit. The dimensions , materials, and technology for each of these reactors are similar. Differences between the B&W units and the Beaver Valley unit are related to di'Lerences in power levels.

Item 3. Preliminary Nuclear Design Data The core size, number of fuel assemblies, and number of fuel rods are the same for all of the B&W units and similar for the Beaver Valley unit. Fuel enrichments differ between the B&W units primarily due to the different fuel cycle burnup requirements.

The excess reactivity requirements for each reactor also vary with fuel cy-cle burnup; the higher burnup of Midland units is reflected in the higher initial excess reactivity. The Midland and Davis-Besse units have fever control rod assemblies than do the Oconee units and more control rods than

_.. the Beaver Valley unit. The reduction in the number of control rod assem-blies and control rod worth for the Midland and Davis-Besse units is due to 19 1-7

D-B less control rod insertion in the core during operation for compensation of equilibrium and transient xenon reactivity changes. The movable control rod worth for shutdown is not changed. The utilization of burnable poison as a part of the control balance allows for a reduction of the soluble poi-son concentration to obtain moderator coefficients within a desired range.

The Doppler coefficient for all cases shown is negative over the core life.

Item k. Principal Design Parameters of the Reactor Coolant System Most of the features in this section are directly related to material prop-erties and the amount of heat produced in the reactor core. Note that the B&W units are identical. The parameters are scaled in proportion to the power of the reactor. The major difference is the number of coolant loops required to remove the heat produced.

For the B&W units, only two loops are required since once-through steam generators are used instead of the U-tubes-in-shell design. The greater l

cooling capacity of these steam generators pemits a reduction in the num-ber of cooling loops for an equivalent amount of heat removed.

Item 5. Reactor Coolent System - Code Requirements The B&W units are identical. Code requirements for the shell side of the steam generator conform to the ASME III Class A Specification. This is con-l sidered to be contribution to the safety of the vessel. It enhances the s integrity because of the more stringent ASME III Class A design, material, l l and quality-control requirements.

l Item 6. Principal Design Parameters of the Reactor Vessel The B&W units are identical. These vessel designs are characterized by a l thinner thermal shield and a relatively larger shell diameter. The larger l diameter provides for additional water between the edge of the core and i the vessel which leads to additional neutron attenuation.

l-( Item 7 Principal Design Features of the Steam Generators I

The steam generators in the B&W units are the same. They are basically dif-ferent from the Beaver Valley unit since they are a once-throug.. lesign in-corporating an integral superheat section.

j Item 8. Principal Design Parameters of the Reactor Coolant Pumps l

e a 8- e8Se Ps are single s W on, single dis d arge pumps, s M e h l3 i

design to those in the Midland and Oconee units.

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l The Davis-Besse head requirements are lower due to the improved piping and '

l -loop arrangement. The B&W reactor pumps have higher head and horsepover

, requirements than the. Beaver Valley unit was for approximately the same N l

flow because of differences in system pressure drops.

20 Amendment No. 3 1-8

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I Item 9. Principal Design Parameters of the Reactor Coolant Piping The B&W designs utilize carbon steel piping clad with stainless steel. The B&W design for Davis-Besse has a new arrange =ent in component elevations . The vessel has been lowered in relation to the steam generato*;. ine vessel is nozzle supported and the steam generators utilize sliding supports to permit thermal expansion of the reactor coolant piping. The improved loop arrangement eliminates the necessity for the internal vent valves utilized in other B&W supplied nuclear steam supply systems. Vent valves were added to these other systems because of the potential for trapping 30 feet of water in the piping at the suction of the pumps following a rupture in the reactor vessel inlet piping. Without the vent valves to provide a path for stean relief, core flooding would be inhibited by 2 the 30 ft water leg. The new arrangement utilized here has the potential for trapping less than 6 feet of water in the pipe. This vater would be blevn out during the blowdown or the core flooding period and core cooling will not be inhibited.

Item 10. Reactor Containment Parameters The containment for the station consists of two structures; one being a free-standing steel containment vessel and the other, a reinforced concrete shield building sur.ounding the vessel. The differences between this and other free-standing steel containments are determined by the design incident te=peratures and pressures which result from station layout, engineered safety features, system capacities, and site location. The shield building design offers satis-factory protection to the surrounding population in case of accident or during normal operation of the station.

Item 11. Engineered Safety Features Engineered safety features are generally similar except for the shield building and penetration roca ventilation and filtration system.

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TABLE l-2 DESIGN PARAMETER $ - DAVl$-BESSE NUCLEAR p0WER STATION Midland Plant Oconne Nuclear Stetten

3. Hydraulic and Tt.ernal Deslan parameters Davis-Besse Beaver Valley Dit t 1 Unit I

. Bated meet Output (core), MWt 2,633 0 2,652 6 2,452 2,568 6 Rated Meat Output (core), 8tu/h 8,987 a 10 9,054 m 10 8,369 m 10' 8.765 m 10 Destga overpower, % 112 112 114 114 System Pressure (nominal), pela 2,200 2,250 2.200 2,200 System Pressure (minimum sesady state), psia 2,150 2,220 2,15e 2,150 Power Distribution factors Heat Generated in Fuel and Cladding, % 97.3 97.4 97.3 97.3 F h (nuclear) 1.75 Not Applicable 1.78 1,78 F (nuc lea r) 2.98 Not Applicable 3.03 3.03 N HothhannelFactors {

F 3.06 3.12 3.12 g N pas b(nuc.

elo atand mech.)

aated Conditions 1.92 (W-3) 1.85 (W-3) 2.21 (W-3) 2.0 (W-3) b Mintenan DNS Ratio at Design Overpower I 50 (W-3) 1.30 (W-3) 1.74 (W-3) 1.55 (W-3)

Coolant Flow 6 6 6 6 Total Flow Bate, Ib/h 138.3 m IO6 100.7sIg 131.3 m 10 6 131.3 m 10 6 EffectiveFlowsateforHeatTransfer,Ib{h 124.2 x 10 96.2 a 10 124.2 m 10 124.2 s 10 7 Effective Flow Area for Heat Transfer, it 49.19 41.8 49.19 49.1g e

Average Velocity Along Fuel Rods, it/s 15.7 14.2 15.7 15.7 0 Coolant Temperature, F Nominal Inlet 557 543.5 555 554 f Average Rise in Vessel 51 67.4 47.8 50.7 Average Rise in Core 52.9 70.2 49.3 51.5 '

Average in Core 583.45 578.6 579.7 579.3 Average in Vessel 582.5 577.2 578.9 578.9 Nominal Outlet of Hot Channel 648.8 644.6 647.1 647.9 Average Film Coef ficient, Btu /h-ft2-F 5,000 5,470 5,000 5,000 Average Film Temperature Dif ference, F 31 38.7 31 31 Heat Transfer at 1001 Power Total Heat Trans fer Surf ace Area, it 49,7 % 42.460 49,7 % 49,734 Average Heat Flux, Bru/h-ft 175,810 207.600 163,725 178,470 Mastmean Heat Flux, stu/h-ft 538,730 543,300 510,300 5 % ,440 Average Tleermal output, kW/ft 5.8 6.7 5.4 5.656 Maximum Thermal Output, kW/ft 17.8 17.9 ,16.83 17.63 i Maatsum Clad Sur f ace Temperature at Nominal Pressure, F 650 657 654 654 Fuel Central Temperature. F Mantenas at 1001 Power at Hot Spot 4,464 - 4,150 4,250 Maximum at overpower 9 4,720 (182%) - 4,450 (114%) 4,650 (184%)

T1.e rnal out put , kW/ft at Mas tanam 19.9 20.0 19.2 20,1 overpower 1

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e T ABLE l-2 (CONT'D)

Midlant Plant Oconne puelear Station Daels-kesse Beaver Vallejt t%tt Dnit

2. Core Mechanteel tw otan Parame' ers Feel Assemblies CRA can; se RCC centess CRA capless CRA centess p.,g g, 0.568 0.563 0.568 0.568 hed Fitch, in.

Overall Dimeestens, to. 8.536 a 8.536 8.426 = 8.426 Total Weight, Ib 8.536 s 8.5}6 8.516 m 8.536 274,350 223,725 274,350 274.350 thumber of Spacer Crfde per Assembly 8 7 8 8 Feet Rode number 36,816 32,028 36,8t6 36,886 Outside Diameter, In. 0.4 30 0.422 0.430 Diametral Cep, to. 0 .4 10 0.007 0.0065 0.007 0.007 Clad Thickness, sn. 0.0265 0.0243 0.0765 Clad Material 0.0265 Etrealey.4 Zircaloy 4 Zi rc a t oy-4 Zircaley-4 Feel Pellets Matettal Density, % of theorectical D02 sintered U02 sintered U02 sintered U02 'I"I"'d 93.5 Region 1 - 94 9 3.5. . . 91.5 Region 2 - 92 Region 3 - 91 Diameter, in. 0.370 0.3669 0.370 0.370 Length in. 0.7 0.6 0.7 0.7 Control Rod Assemblies (CRA)

pentron Absorber 5% Cd-t5% In-80% Ag 8C g,

4 5% Cd-15% In-BDI Ag 5% Cd-15% In-80% Ag &

B3 Length of Folson Sectioe, in. 1 34 144.25 1 34 Cladding Material 1 34 304 $$, cold worked 304, SS cold worked 304 SS, cold worked 304 $$, cold worked Clad Thickness, in. 0.021 0.0t95 pumber of Assemblies 0.021 0.021 49 45 49 61 Ihamber of Control Rods per Assembly 16 20 16 16 Antal Power Shaping Rod Assemblies (AFSRA) poutron Absorber 5% Cd,15% In 8DI Ag BC4 5% Cd 15% In,80% Ag 5% Cd,85% In.80% As Length of Poison Section, in. 36 36 36 36 Cladding Material (Poison Section) 304 SS, cold worked 3n4 SS, cold worked 304 SS, cold worked Cisd Thickness, in.

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304 $$, cold worked 0.021 0.0*95 0.021 0.021 Ikssber of Assemblies 8 8 8 8 theber of Control Rods per Assembly I6 20 16 16 Burnable Poison Rod Assemblies (BPRA) pg ) peutron Absorber Al 234 0 9 C Borosilicate glass Al go3 49C Length of Poison Section, In.

144 142.68 144 l P4 Clodding Matertal Clodding Thickness,ta.

Zircalay-4,ecid worked Stainless Steel Zircalay-4, cold worked

.4 35 .0 15 pueber of Assemblies 72 68 72 thesber of Reds per Assembly 16 12 16 Orf f tee Rod Assemblies (ORA)

Rod Material 304 $$, annealed 304 SS 304 $$, asncated pueber of Ortitce Rods per Assembly 304 SS, annealed 16 verfes 16 16 Core Structure Core Barrell 1D/0D, in. 14t/145 133.9/137.9 141/t45 Thermal Shield ID/00 in. 141/145 147/851 144.1/149.5 147/l51 147/I51 N .

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.. w TAAIE I-2 (CONT'D)

Daw l s - Be s s e Beaver Valley MIJiant Plant Gronee Nuclear Station .

Unit I Unit t

3. Preliminary Nuc lea r nes t yn pa t . 'i e' Structural chasarteristics Fuel Weight ( as U02)e Ab. 207,486 176,200 44a4 Weight (ar tive zone), Ib 207,466 207,466 42,200 36 ,300 42,200 42,200 rure Diameter.ine. (equivalent) 128.9 189.5 428.9 128.9 Core Height, I nc . ( ac t ive field) 144 144 g44 144 keflector Thickness and Composition

Top (water plus steet), in. 12 to 12 12

, bottom (water plus stee t), in. 12 to 12 12 Side (Water plus steel), in. 18 15 IS 18 Number of Fuel Assemblies 477 157 177 177 N Fuel Rod /Fust Assembly 208 204 208 200 Performance Characteristics

-E2h Loading Technique 3 region 3 region 3 region 3 region Fuel Discharge 3ernup, Mwd /MTU 13,686 I4,500 13,540 9,600 (Average-First Cycle)

Feed Enrictaments, wt 10-235 (First Cycle)

- Region i 2.32 2.0 2. 30

.- 2.10 a Region 2 2.32 2.6 2.30 J.

Region 3 avs 2.68 3.2 2.64 Cont rol Clearacteristics Ef fective Mult iplication (Beginning of Life)

ColJ, No Pawer, Clean 4.276 1.207 1.273 1.248 Not, No Power, Clean I.204 t.t63 1.218 1.198 Hoe , Rated Power Xe Equilibrium 1.088 B.137 t 084 4. 8 34 Control Rod Worth gik/h), t 8.0 8.16 8.0 10.6 Boron Concentrations To Shus Reactor Down With All Rods Inser tcJ (clean), colJ/ hot ppm 836/484 880/6t0 810/450 664/587 8oron Wurth (but),1 gik/h), ppm 1/100 1/85 1/100 1/85 Boron Worth (cold),1 fik/h), ppm I/75 1/70 t/75 t/64 Rinet te Charac teristics (Range During Life Cycle) ,4 Ma4eratut Temperature Coeflicient, gik/h) /F 0 ' O to -3,0 m 10' 4 to.3 t 10 Heterator l'ressure Coef fic icut . 61k/k) / psi 64xIO~go-3.0m1010-6 to + 3.0 a -0.3 m 10'g -3.5 m Ibto +3.5 x 10-6 to -3.0 a 10

Moderator Void Coef ficient , gik/k) 1 void -4.0 x 10-4 to +3.0 m 10*3 -2.5 x 10-3 t o +0.5 x 10* 3 +4.0 x 10*'7

+4.0 a 10* to -3,0to +3.0 x 10*I a 10-6 to +3.0 m

+0.5 to 10' 1.6 s x-5.0 10*3 m 10'I Doppler Coef ficient , 61k/h) /F al.1 x 10*I to I.7 x 80-5 .l.0 x 10*5 to 't.7 m 10-5 .g,g ,30-5 to l.7 m 10-5l.1+4.0 x 10-5 mg.10-4 3,y , 30-5 4 Principal th sign Paramriers of t he Reactor Coolan t System System Heat (ht put , MWt 2,650 2,652 6 2e468 6 2.584 6 System reat output, stu/h 9,044 x IO* 9,058 x 10 8,423 m to 8,849 x 10 Operatin. Pressure, psig 2,185 2,235 2,I85 2.185 Reactor Inlet Temperature, F 557 543.5 555 5 54 Reactor Outlet Temperature, F 608 610.9 603 6 04 Number of Loups 2 3 2 2 Design Pressure, psip 2,500 2.485 2,500 2,500 Design Temperature, F 650 650 650 650 Hydrostatic Test Pressure (cold), Psig 3,125 3,110 3.125 3.125 3

Coolant volume, inc luding pressurizer, it 18,440 9.000 ft,800 11,478 Total Reactor Flow, gym 352,000 265,500 352,000 352,000 t

_m_____ _ _ _ _ _ _ _ _ _ _ _

h..

TABLE l-2 (C0KT'D)

?tidiant Plant pronce Nuclest Statlan Davis-heese & aver Valley t a t t _t Dnit 8

5. mesctor Coelant System Code Reestrements Beacter Vessel and Closure Need - ASME !!!, Class A ASFE lit. Class A ASME lil, Class A ASME Ill, Class A-Steam Generater

, Tube Slee ASME III, Class A ' ASME III, Class A ' ASNE III,fless A AtME lit,Cla<s A Shell tide ASME III, Class A 45MF III, fla<e A A%HF III, Class A ASME fil, Class A Presseels e ASMF III, Class A At9[ lit , Class A ASMF lit, Class A AKMF Ill,(1a54 A Pressuriser Safety valves ' ASME Ill, Art. 9 ASME Ill. Class A A5ME III, Art. 9 A9MC Ill, Art.9 geector Coctant Fiping ISASI 238.7 05A11 til.1 t5Asl 331.2 45ASI B31.7 Scoctor Cootent Pump Casing A tME III, Class A ---

AsMr Ill, Claes A AtMF lit, Class A Inot fede Stamqvdl Innt rode Stamped) Imot (pdc Stamped)

6. Frf ncipal Deslan Farameters of the Beector Vessel Material %A- s t), Cr ade B, glad with 5 A- 902 cr ede p %A-5 ,,,eiede e, clad SA-513. Crede 8, clad 18-p Kr aintee s steel etad with 70*. wit h 18-8 Statalese with 18-8 stainlese Stainless se cc i srcel

$ rect 2,500 2,500 '

[

C7 Design Fressure, psig

. Desira Temperature, F. 650 2,585 6'0 650 2,500 650 {

Operat ing Fressure, psig 2,88% 7, 2 35 2,l85 2,885 Ing s de Di eme n , " o f She ll in. 171 157 179 Ill Out side Diameter across Norries in. 29 252 -249 269 Overall nelaht of Ve**el and Clasere Ik-ad (over Cpp and inst rument. partles),

It.in. 79/n $2/7,Isa 60/8-3/4 60/8-3/4 Minimum Clad Thic6eces,in. I/F 0.156 1/8 ' I/R 2, Princ ipal Desirn Farnes ters of the Steam Generatorg

]'

. Number o f I ni t s 2 1 2 2 Type rrr e ir al, pace -thron,-h ve rt ic al r-tube with Yrrt6 cal, once-through Vert ical, once.thenerb wit h ins crral .nperkrat, r laterral mas. tere with lategral superheater with integral superheatee orparator

[5gjp Tube Material las pac t Inc one l smenact Inc ene t shell Material e Jr ban St et I (Jfbun Steel ( Arbon $ teel Carbon StrrI k II - Tube Side Dreign Pressure, pair. 2,ian 2,6R) 2,500 2,500 Tube Side Desirn Temperature, F $$n 650 650 6%o Tube Destra Flow, th/h 6 v.f 4 x In* 31.57 x 10 65.64 s 10' 65.66 s In ShcIl Side Dreign Pregsvre, rels I.nio I,ns , 1,050 I,050

' shcIl Side Desien Tcoperature, F 600 600 6no 600 operating Pressure, T ne sgge i

unmanal, psia 2,Isi 2,2 35 2,885 2,IS)

Operating Pressure, Shell tide, poninal, pelg 990 8,ntis 910 98n Supe r heat at outlet at Rated Lead 15F --- )$F 35F Nydrestat ic Test Pressure 8 tobe-side cold),

psig 1,125 1,806 3,825 3,825

/

)

a J

1 s

i s TA8tf I-2 (CONT'D)

Mid las.t Plant Oconee Nuclear Station Daviu-hesse Seaver Valley Unit I Unit I

8. . Principal Design Parameters of the Reactor Coolant Ptsope

, Number of Units 4 3 6 Type 6 Vertical, sir.gle stage . Vertical, single stage Vertical, single stage 8

,-Design Pressure, psig Vertical, single stage 1 2,500 2,485 2,500 2,500

, Design Temperature P 650 650 650 650 Operating Pressuse, Nominal, psig 2,185 2,235 2,185 2,I85 Suction Temperature. F 557 543.3 554 Design Capacity, gpm 554 88,000 88,500 , 88,000 88,000 N Design Total Deve loped HeaJ, it. 355 280 370 396 Hydrostatic Test Pressure (cold), psig 3,125 3,106 3.125 3,750

@ Motor Type . a-c induc tion, single a-c induct ion, single a-c induction, slagte a-c induction, single speed speed speed speed Motor Rating (nameplate), hp 9,000 6,000 9,000 9,000 9 . Principal Design Parameters of the Reactor Coolant Pipinn Miterial Carbon steel claJ with SS Hot isg ( 50), in.

Austenitic SS Carton steel clad with SS Carbon steel clad witti $$

36 29 36 36 Cold 1sg (ID), in. 28 27.5 28 28 Between Psap and Ste.ua Generator (ID),in. 28 10 Containment system Parameters 31 28 28

{

Type t ree standing steel contain- Rainf orced concrete,

.nent vessel of SA-299 mat'l Steel-lined, prestressed, Steel-lined, prestressed.

3 ag,,3. lined with flat with cylindrical walls, hemi* post-tensi ned concrete, post-tensier.ed concrete, base, cylinder walla vertical cylinder with vertical cylinder with spl.erical head and ellipsoi- and hemispleerical done Jal bot tum encl. by rein. con- flat bott<ss and shallow flat bottoss and shallow Design Par 4eacters rete hield build g. daomed roof. Joomed roof, Inside Diameter, it conta nment Vesse 130 126 116 IM Height, it.

280 185 (Scaled) 193 206 Frce Volume, It I 2,2l00,000 lesign Pressure, psig (1)

I 800,000 I.670,000 I 900,000 40 42.5 58 59 Concrete Thickness , It (Shield Building)

Vertical Wall 2 4h U p., 3*3/4 gg 25 3 Containment Isak Prevention 3l and Mitigation leak-tight penetrations, leak-tight penetrations leak-tight penetrations c on t aisumen t vessel and leak-tight penetrations aW continues steel liner, ad continuws steel !!aer, shielJ butIJing, and continuous steel liner. Auromatic isolation where Automatic isolatfor,where Automatic Isolation where required. required.

required, Normal opera-tion is at sub-atmospneric .i j

pressure.

j Caseous Ef8luent Purge Discharge vent above Discharge went line 10 f t . Discharge vent line at top Discharge sent line at top shield building. above turbine building roof, of reactor building. of reactor building.

(1) 'hanimian internal pressure *' as defined in Article Nf 3 81 of ASME Code,Section III.

--_ -. _._._-___----..__.-_.-____-___-_.) -

--l

Y d

TAB 12 1-2 (CONT'D) 4 Nidlant Plant Ocomee paclear Station Dawls-Besse Beaver Valley Unit i Unit I ill, ' Engineered Safety Features hoergency _ injection Syetem shamber of fligh Head Pumpo 3 3 leumber of Im Head Pumps 3 2 2

'3 Contaltunent Coolers -

2 2 thenber of finits -3 3 5

4 3.

Air Flor Capacity, Each, at Accident 58,000 Condition, cfm 100,000 21,000 Core Flooding System 54,000

. Isumber of tanks 2

Total Voliene, Each, f t3 '3 2 Beactor Building Spray I 410 1,450 2 1,430 ' 3,430 letsaber of hoops Emergency Power 2' 2 2

Cenerator tintts, No. Type 2

2. Diesel 2, Diesel 2 for both units, Diesel s

. Isot comparableI*)

Y E .=-

4' ten O s 70-MW hydroelectric units with one overhead and one underground feeder. Also, one of three 44 +tv4 gas turbine units located 30 miles distant dedicated solely for backup emergency power.

4 N

l 1

v -. v

D-B 1.h PRINCIPAL DESIGN CRITERIA The principal design criteria for Davis-Besse Unit 1, were developed in con-sideration of the 70 General Design Criteria for Nuclear Power Plant Con-struction Permits proposed by the AEC in its press release, K-172, of July 10, 1967. Listed below are the 70 criteria proposed by the AEC, together with applicant's response indicating, with respect to each criterion, the principal architectural and engineering criteria adopted by the applicant.

In the discussion of each criterian, reference is made to sections of the report where more detailed infctmation is presented.

1.k.1 CRITERION 1 - QUALITY STANDARDS (Category A)

Those systems and components of reactor facilities which are essential to the prevention of accidents which could affect the public health sad safety or to mitigation of their consequences shall be identified and then designed, fabricated, and erected to quality standards that reflect the importance of the safety function to be performed. Where generally recognized codes or standards on design, materials, fabrication, and inspection are used, they -

shall be identified. Where adherence to such codes or standards does not suffice to assure a quality product in keeping with the safety function, they shall be supplemented or modified as necessary. Quality assurance pro-grams, test procedures, and inspection acceptance levels to be used shall be identified. A showing of sufficiency and applicability of codes , stan-dards, quality assurance programs, test procedures, and inspection accep-tance levels used is required.

Discussion

a. Essential Systems and Components The integrity of eystems , structures , and components essential to accident prevention and to mitigation of accident consequences has been included in the reactor design evaluations. These sys-tems, structures, and components are:

^1. Reactor coolant system.

2. Reactor vessel internals.

3. Containment vessel.

h. Engineered safety features.

5 Electric emergency power sources.

b. Codes and Standards The following table references applicable sections where codes, quality control, and testing are included in the PSAR. The Quality Assurance program is discussed in detail in Appendix 1B.

28 1-17

D-B Quality Test Item Codes Control Procedures s

)

a.1 h.1.5 5.5 h.h a.2 3.1.2.h.1, 3.2.h.1 3.1.2.h.1 3 3.h a.3 5 1.1 5.2.2.h 5 2.1.h 5.8.1 5.8.1 5.2.1.h 5.8.2 5.8.2 5.2.2.h App 5A App 5C App SC App 5B App SC a.h 9(p 9-1) 6.1.4, 6.2.h, 6.3.h a.5 8.1 8.3 136 1.k.2 CRITERION 2 - PERFORMANCE STANDARDS (Category A)

Those systems and components of reactor facilities which are essential to the prevention of accidents which could affect the public health and safety or to mitigation of their consequences shall be designed, fabricated, and erected to performance standards that vill enable the facility to withstand, without loss of the capability to protect the public, the additional forces that might be imposed by natural phenomena such as earthquakes , tornadoes ,

flooding conditions, vinds, ice, and other local site effects. The design bases so established shall reflect: (a) appropriate consideration of the most severe of these natural phenomena that have been recorded for the site and the surrounding area and, (b) an appropriate margin for withstanding )

forces greater than those recorded to reflect uncertainties about the his-torical data and their suitability as a basis for design.

Dis cussion

a. Essential Systems and Components The integrity of systems, structures, and components essential to

. eaccident prevention and to citigation of accident censequences has been included in the resctor design evaluations. These syt -

ter.s, structures, and components are:

1. Reactor coolant system.

2. Reactor vessel internals.

3 Containment vessel.

h. Engineered safety features.

5 Electric emergency power sources.

)

1-18 29

_~

D-B

b. Performance Standards s

These essential systems and components have been designed, fabri-cated, and erected to performance standards that will enable the facility to withstand, without 1 css of the capability to protect the public, the additional forces that might he imposed by natural phenomena. The designs are based upon the most severe of the natural phenomena recorded for the vicinity of the site, with an appropriate margin to account for uncertainties in the historical data.

These natural phenomena are listed below. The design bases for these forces are included in Appendix 5A of the PSAR.

1. Eerthquake

2. Tornado 3 Grcund Water and Flood *

h. Wind and Hurricane 5 Snow and Ice

6. Other Local Site Effects 1.k.3 CRITERICU 3 - FIRE PROTECTION (Cate6ory A)

The reactor facility shall be designed (1) to minimize the probability of events such as fires and explosions and (2) to minimize the potential effecte of such events to safety. Noncombustible and fire resistant materials shall be used whenever practical throu6 hout the facility, particularly in areas containing critical portions of the facility such as containment, control room, and components of engineered safety features.

Discussion Ncncombustible and fire resistant materials are used whenever practical through-out the facility, particularly in areas containing critical portions of the station such as containment structure, control room and components of the engi-neered safety features systems.

Equipment and facilities for fire protection (9.11), including detection, alarm and extinguishment are provided to protect both station and personnel from fire, explosion and the resultant release of toxic vepors. Both wet and dry type fire-fighting equipment will be provided.

Normal fire protection is provided by deluge systems, sprinklers, hose' lines and portable extinguishers.

The fire protection system will be designed in accordance with the requirements of the Nuclear Energy Property Insurance Association as a guide and applicable codes and regulations of the State of Ohio.

The fire protection system will be provided with test hose valves for periodic testing. All equipment is accessible for periodic inspection.

s-

^ '

30 1-19

D-B 1.L.L CRITERION h - SHARING OF SYSTEMS (Category A)

Peactor facilities shall not share systens or components unless it is shown i

safety is not impaired bp the sharing.

Discussion Davis-Besse is a single unit plant and this criterion is not applicable.

l.h.5 ' CRITERION 5 - RECORD REQUIREMENTS (Category A)

Records of the design, fabrication, and construction of essential components of the plant shall be maintained by the reactor operator or under Jts control throughout the life of the reactor.

Discussion After the construction of the station has been completed, records, correspondence, field books, forms , film, temperature charts and other supporting data required to show compliance with the codes, tests and quality control standards as out-lined in the PSAR, vill be in the possession of or available to The Toledo Edison Company throughout the life of the statior (Appendix 1B).

1.L.6 CRITERION 6 - REACTOR CORE DESIGN (Category A)

The reactor core shall be desiFned to function throughout its design life-time without exceeding acceptable fuel damage limits which have been stip-ulated and justified. The core design, together with reliab) process and decay heat removal systens, shall provide for this capability under all ex- -)

pected conditions of normal operation with appropr ste margins for uncer-tainties and for transient situations which can be anticipated, including the effects of the loss of power to recirculation pumps, tripping out of a turbine generator set, isolation of the reactor from its primary heat sink, and loss of all off-site power.

Discussion

. The reactor is designed with the necessary margins to accommodate, without fuel damage, expected transients from steady-state operation including the transients given in the criterion. Fuel clad integrity is ensured under all normal and abnormal modes of anticipated operation by avoiding clad overstressing and overheating. The evaluation of clad stresses includes the effects of internal and external pressures, temperature gradients and changes , clad-fuel interactions , vibrations, and earthquake effects. The free-standing clad design prevents collapse at the end volume region of the fuel rod and provides sufficient radial and end void volume to accommodate -

clad-fuel interactions and internal gas pressures (3.2.h.2).

l Clad overheating is prevented by satisfying the core thermal and hydraulic l . criteria (3.1.2.3 and 3.2.3.1.1):

t a

\w-1-20

                                                                               -- m .-  _,_ m, D-E

e. At the design overpower, no fuel melting vill occur.

b. A 99 ner cent confidence exists that (t least 99 5 per cent of the f ael rods in the core will be in no jeopardy of experiencing a DNB daring continuous operation at' the design overpower of 112 per cent (1.12 x 2633).

The design margins allow for deviations of temperature, pressure, flow, re-actor power, and reactor-turbine power mismatch. Above 15 per cent power, the reacto'r is operated at a constant average coolant temperature and has a negative power coefficient to damp the effects of power transients. The reactor control system will maintain the reactor operating parameters within

        - preset limits, and the reactor protection system will shut down the reactor if t ormal operating limits are exceeded by preset amounts (7.1 and ik).

Reactor decay' heat will be removed through the steam generators until the reactor coolant system is cooled to 280 F. Steam generated by decay heat will supply the steam-driven main feedwater pump turbines, and can also be vented to atmosphere and/or bypassed to the condenser. The steam gener-ators are supplied feedwater from either the main steam-driven feedwater pumps, or from a turbine-driven emergency feed pump, sized at 3 per cent of full feedwater flow. The main feedwater pumps supply the steam generators with water contained in the feedwater train and the condensate storage tank. The emergency feed pump takes suction from the condensate storage tank. These sources provide at least 500,000 gallons of water storage, which is sufficient to remove

 <       decay heat for about one day after reactor shutdown with the primary heat sink (condenser) isolated. The condenser and deareator are normally avail-able so that water inventory is not depleted (10.2.4), even in the event loss of electrical power.

The reactor coolant pumps are provided with sufficient inertia to maintain

       - adequate flow to prevent fuel damage if power to all pumps is lost. Natural circulation coolant flow will provide adequate core cooling after the pump energy has been dissipated (1h.1.2.6).

l.h.7 CRITERION 7 - SUPPRESSION OF POWER OSCILLATIONS (Category B) The core design, together with reliable controls, shall ensure that power oscillations which could cause damage in excess of acceptable fuel damage

       - limits are not possible or can be readily suppressed.

Discussion Power oscillations resulting from variations of coolant temperature are minimized by constant average coolant temperature when the reactor is op-ersted above 15 per cent power. Power oscillations from spatial xenon ef- < fects are minimized by the large negative power coefficient and axial rower

       - shaping rod assemblies. Reactor overpower trip prevents fuel-clad damage resulting from DNB.

3 2-1-21 m

D-B Tha ability of the reactor control and protection system to control the os-cillations resulting from variaticn of coolant temperature within the con-tral system dead band and from spatial xenon oscillations has been analyzed. ) Variations in average coolant temperature provide negative feedback and en-hance reactor stability during that portion of core life in which the mod-orator temperature coefficient is negative. When the moderator temperature co3fficient is positive, rod motion vill compensate for the positive feed-bock. 'lhe maximum rate of power change resulting from temperature oscilla-tions within the control system dead band has been calculated to be less tnan 1 per cent / minute. Since the unit has been designed to follow ramp lond changes of 5 per cent / minute, this is well within the capability of the control system. Control flexibility, with respect to xenon transients, is provided by the combination of control rods and nuclear instrumentation. Axial, radial, or czimuthal neutron flux changes vill be detected by the nuclear instrumenta-tion. Indivi$2al control rods or groups of control rods can be positioned' to suppress and/or correct flux changes. The analysis of xenon-related power effects is presented in BAW-10010, " Stability Margin for Xenon Oscil-lations - Modal Analysis." 1.k.8 CRITERION 8 - OVERALL POWER COEFFICIENT (Category B) The reactor shall be designed so that the overal power coefficient in the power operating range shall not be positive. Discussion Th2 overall power coefficient is negative in the power operating range (3.2.2). 1.k.9 CRITERION 9 - REACTOR COOLANT PRESSURE BOUNDARY (Category A) The reactor coolant pressure boundary shall be designed and constructed so as to have an excaedingly low probability of gross ruptura or significant leakage throughout its design lifetime. . Discussion Tha reactor ':oolant system pressure boundary meets the criterion through the following:

a. Material selection, design, fabrication, inspection, testing, and certification in accordance with ASME codes.

b. Manufacture and erection in accordance with approved procedures,

c. Inspection in accordance with ASME code requirements plus addi-tional requirements imposed by the manufact.urer.

d. System analysis to account for cyclic effects of thermal tran-sients, mechanical shock, seismic loadings, and vibratory loadings.

e. Selection of reactor vessel material properties to give due con-sideration to neutron flux effects and the resultant increase of ,-

the nil-ductility transition temperature. 1-22 )

D-B

  ,-s The materials, codes, cyclic loadings, and non-destructive testing are dis-cussed further in Secti,on h.

1.4.10 CRITERION 10 - CONTAINMENT (Category A) Containment shall be provided. The containment structure shall be designed to sustain the initial effects of gross equipment failures, such as a large coolant boundary break, without loss of required integrity and, together with other engineered safety features as may be necessary, to retain for as long as the situation requires the functional capability to protect the public. Discussion The containment vessel' (Section 5.1) and the engineered safety features (Sec-tion 6) are designed to prevent undue risk to the health and safety of the public from the consequences of an unlikely event of loss-of-coolant accident (Section Ih.2.2.3), which is based on a postulated break of the reactor coolant , piping up to and including a double ended break of the largest reactor coolant pipe. The containment vessel and the engineered safety features are designed to safely sustain all internal and external loading conditions that may reason-ably be expected to occur during the life of the station, including a loss-of-coolant accident. Due consideration has been given to all site factors and local environment as they relate to the public's health and safety. 1.k.11 CRITERION 11 - CONTROL ROOM (Category B) The facility shall be provided with a control room from wich actions to maintain safe operational status of the plant can be controlled. Adequate radiation protection shall be provided to permit access, even Under accident conditions, to equipment in the control room or other areas as necessary to shut down and maintain safe control of the facility without radiation exposures of personnel in excess of 10 CFR 20 limits. It shall be possible to shut the reactor down and maintain it in a safe condition if access to the control room is lost due to fire or other cause. Discussion Following proven power station design philosophy, all control stations, switches, controllers, and indicators necessary to start up, oyerate and shut down the nuclear unit and maintain safe control of the facility will be located in one control room (T.5). Safe occupancy of the control room during abnormal conditions will be provided for in the design of the control room. Adequate shielding vill be used to main-tain tolerable radiation levels in the control room for hypothetical accident < conditions. A positive control room pressure vill be maintained to prevent in-leakage. The control room ventilation system will be provided with radiation detectors and appropriate alarms. Provisions will be made for control room air to be recirculated with mintamn makeup through high efficiency char- 8 coal filters. Emergency lighting vill be provided. 1-23 Amendment No. 8

}f

NB

Controls remote from the main control room vill be available to bring and main-tain the reactor at a hot standby condition in the event access to the main con- ) trol room is lost due to fire or other causes. 1.h.12 CRITERION 12 - INSTRUMENTATION AND CONTROL SYSTEMS (Category B) Instrumentation and controls shall be provided as required to monitor and maintain variables within prescribed operating ranges. Discussion Adequate instrumentation and controls are provided to maintain operating variables within their prescribed ranges, including nuclear instrumentation, incore instrumentation, nonnuclear instrumentation, and an integrated control system. The nuclear instrumentation monitors reactor power from source level to 125 per cent full power (7.3.1). The incere instrumentation monitors power dis-tribution throughout the reactor core (7.3.3). The nonnuclear instrumenta-tion measures temperatures, pressures, flows, and levels in the reactor cool-ant system, steam system, and auxiliary reactor systems, and maintains these variables within prescribed limits (7.3.2). The integrated control syste= matches reactor, steam generator, and turbine load to electrical demand and maintains constant average reactor coolant tem-perature from 15 to 100 per cent of rated load and constant steam pressure at all loads (7.2).

1.h.13 CRITERION 13 - FISSION PROCESS MONITORS AND CONTROLS (Category B) Means shall be provided for monitoring and maintaining control over the fis-sion process throughout core life and for all conditions that can reasonably be anticipated to cause variations in reactivity of the core, such as indica-tion of position of control rods and concentration of soluble reactivity con-trol poisons. Discussion This criterion is met by reactivity control means and control room display. Reactivity control is by movable control rods and by chemical neutron ab-sorber (in the form of boric acid), dissolved in the reactor coolant. The position of each control rod will be displayed in the control rcom. Changes in the reacti'vity status due to soluble boron vill be indicated by changes in the position of the control rods. Actual boron concentration in the re-actor coolant is determined periodically by sampling and analysis (7.3.2.1). 1.b.1h CRITERION lh - CORE PROTECTION SYSTEMS (Category B) Core protection systems, together with associated equipment, shall be de-signed to act automatically to prevent or to suppress conditions that could result in exceeding acceptable fuel dsmage limits. Discussion The reactor design meets this criterion by reactor trip provisions and s-engineered safety features. The reactor protection system is designed to 1-2h 35 . e

e

   .                                               D-B s

limit reactor power which might result from unexpected reactivity changes, and provides an autenatic reactor trip to prevent exceeding acceptable fuel damage limits. In a loss-of-coolant accident, the safety features actuation system automatically actuates the high-pressure and low-pressure injection systems. The core flooding tanks are self-actuating. Certain long-term operations in the emergency core cooling systems which do not require im-mediate actuation are performed manually by the operator, such as remote switching of the low-pressure injection pumps to the recirculation mode and sampling of the recirculated coolant. l.h.15 CRITERION 15 - ENGINEERED SAFETY FEATURES PROTECTION SYSTEMS (Category B) Protection systems shall be provided for sensing accident situations and initiating the operation of necessary engineered safety features. Discussion The Safety Features Actuation System senses reactor coolant system pressure and containment vessel pressure and initiates emergency core cooling, isola-tion and containment vessel cooling at the appropriate level.' l1 l 1.k.16 CRITERION 16 - MONITORING REACTOR COOLANT PRESSURE BOUNDARY l (CateEory B) ' Means shall be provided for monitoring the reactor coolant pressure boundary to detect leakage. Discussion Reactor coolant pressure boundary integrity can be continuously monitored in the control room by surveillance of variation from normal conditions for the following: a Containment vessel temperature, humidity and accumulation of cooling coil condensate.

b. Containment vessel radioactivity levels.

c. Condenser off-gas radioactivity levels (to detect steam gener-ator tube leakage).

d. Increasing purification system letdown storage makeup (indicating system leakage).

Gross leakage from the reactor coolant boundary vill also be indicated by a 1 decrease in pressurizer water level and a large flow from the containment ! vessel sump (h.2.7). b b. 1-25

                                                      ,           y           -

3 D-B

                                                                                      's' 1.k.17        CRITERION 17 - MONITORING RADIOACTIVITY RELEASES (Category B)

Means shall be provided for monitoring the containment atmosphere, the facility effluent discharge paths, and the facility environs for radioactivity that could be released from normal operations, from anticipated transients, and from accident conditions. Discussion Means are provided for monitoring the containment vessel atmosphere, the facility effluent discharge paths, and the facility environs for radio-activity that could be released from normal operations, from anticipated transients and from any accident (11.2.3). Sources of normal radioactive discharges are the liquid and gaseous vastes. The major source of liquid wastes are the bleed-off of reactor coolant from 1l the makeup and purification system and bleeding of reactor coolant prior to refueling. The release path of gaseous vastes is through system vents. The gases are collected in the vaste gas surge tank, then compressed. The gases may be released to the atmosphere or stored in vaste gas decay tanks. The condenser air ejector discharge vill be monitored for gaseous activity. l.h.18 CRITERION 18 - MONITORING FUZL AND WASTE STORAGE (Category B) T

Monitoring and alarm instrumentation shall be provided for fuel and vaste storage and associated handling areas for conditions that might result in loss of capability to remove decay heat and to detect excessive radiation levels. Discussion Monitor:Lg and alarm instrumentation is provided sensitive to the operation of the spend fuel pool cooling system (9.4). Heat is removed from stored radvaste by conduction to the ventilation air. Ventilation air from the rad-waste facility is continuously monitored for radioactivity (11.2.3.2). l.h.19 CRITERION 19 - PROTECTION SYSTD(S RELIABILITY (Category B) Protection systems shall be designed for high functional reliability and in-service testability commensurate with the safety functions to be performed. Discussion The protection systems design meets this criterion by specific instrument location, component redundancy, and in'-service testing capability. The major design criteric stated below have been applied to the design of the instru-mentation. I u-37 1-26

                                                                              . - - ~  .~

D-B

,7 f    9

a. No single 7ponent failure shall prevent the protection systems from ful" '. ling their protective function when action is required.

b. No single component failure shall initiate unnecessary protection syster sction, provided implementation does not conflict with the cr' > ..on above.

Test connections and capabilities are built into the protection systems to provide for:

a. Pre-operational testing to give assurance that the protection sys-tema can fulfill their required functions.

b. On-line testing to assure availability and operability. (7.1.1) 1.h.20 CRITERION 20 - PROTECTION SYSTEMS REDUNDANCY AND INDEPENDENCE (Category B)

Redundancy and independence designed into protection systems shall be suf-ficient to assure that no single failure or removal from service of any com-ponent or channel of a system will result in loss of the protection function. The redundancy provided shall include, as a minimum, two channels of protec-tion for each protection function to be served. Different principles shall be used, where necessary, to achieve true independence of redundant instru-mentation components. Discussion Reactor protection is by fear channels with 2/h coincidence, and engineered safety features are by sour channels with 2/h coincidence. All protection system functions are imp. emented by redundant sensors, instrument strings, logic, and action devices that combine to form the protection channels. Redundant protection channels and their associated elements are electrically independent and packaged to provide physical separation. The reactor pro-tection system initiates a trip of the channel involved when =odules or equipment is removed (7.1.1). 1.h.21 CRITERION 21 - SINGLE FAILURE DEFINITION (Critagory B) Multiple failures resulting from a s!ngle event shall be treated as a single j failure. Discussion j I The crotection systems meet this criterion 10 that the instrumentation is j designed so that a single event cannot result in multiple failures that would prevent the re, quired protective action (7.1.3). l %d 38 E-ST

D-B _ 1.h.22 CRITERION 22 - SEPARATION OF PROTECTION AND CONTROL s INSTRUMENTATION SYSTEMS (Category B) ) Protection systems shall be separated fran c utrol instrumentation systems to the extent that failure or removal from serv. e of any control instrumentation system component or channel, or of those coma 1 to control instrumentation and protection circuitry, leaves intact a system stisfying all requirements for the protection channels.- Discussion The protection systems ' instrument strings are electrically and physically in-dependent. Shared instrumentation for protection and control functions satis-fies the single failure criteria by the employment of isolation techniques to the multiple outputs of various instrument strings (7.1.3.5) . 1.h.23 CRITERION 23 - PROTECTION AGAINST MULTIPLE DISABILITY FOR PROTECTION SYSTEMS (Category B) The effects of adverse conditions to which redundant channels or protection systems might be exposed in common, either under normal conditions or those of an accident, shall not result in a loss of the protection function. Discussion The protective systems are designed for continuous operation under adverse environments. The reactor protection system instrumentation within the contain-ment vessel is designed for continuous operation in an environment of 40 to 120 F < and 90 per cent relative humidity. The neutron detectors are designed for con-tinuous operation in an environment of 175 F, 90 per cent relative humidity, and 60 psig. Engineered safety features equipment and vital instrumentation inside the reactor building are capable of performing required safety functions for conditions h0 psig, 26h F, and 100 per cent RH), which are in excess of the

@   requirements of the loss-of-coolant accident (7.1.1 ar4 7.1.2).

The protection systems ' instrumentation will be subject to environmental (quali-fication) testing as required by the proposed IEEE Criteria for Nuclear Power Plant Protection Systems, IEEE 279, August 30, 1968. Protective equipment out-side the containment vessel, control room, and relay room is designed for con-tinuous operation in an ambient temperature of 110 F and for 80 per cent relative humidity. (71.1h) 1.h.2h CRITERION 2h - EMERGENCY POWER FOR PROTECTION SYSTEMS (Category B) In the event of loss of all offsite power, sufficient alternate sources of power shall be provided to permit the required functioning of the protection systems. Discussion In the event of loss of all offsite power, station batteries and diesel gene-i rators will provide power to permit the required functioning of the protection systems (8.2.3). 39 [ Amendment.No.~3 1-28

D-B l.k.25 CRITERION 25 - DEMONSTRATION OF FUNCTIONAL OPERABILITY OF

 ~/ i                  PROTECTION SYSTEMS (Categorf B)

Means shall be included for testing protection systems while the reactor is in operation to demonstrate that no failure or loss of redundancy has occurred. Discussion Test circuits _ are supplied which utilize the protection systems redundancy, independence, and coincidence features. This makes it possible to manually initiate on-line trip signals in any single protection channel in order to test irip espability in each channel without affecting the other channels (7.1.3). l.h.26 CRITERION 26 - PROTECTION SYSTEMS FAIL-SAFE DESIGN (Category B) The protection systems shall be designed to fail into a safe state or into a

state established as tolerable on a defined bases if conditions such as disconnection of the system, loss of energy (e.g. , electric power, instrument air), or adverse environments (e.g., extreme heat or cold, fire, steam, or water), are experienced.

Discussion , The reactor protection system will trip the reactor on loss of power. The engi-s neered safety features are supplied with multiple sources of electric power for control and valve action. A cotal loss of electrical power to the engineered safety features actuation system vill cause it to assume a tripped position with the exception of the control relays. These relays require power to trip. How-ever, since the engineered safety features equipment also requires power to op-erate, this relay need not assume the tripped position upon a total loss of power. The system is designed for continuous operation under adverse environments. The reactor protection system instrumentation within the containment vessel is designed for continuous operation in an environment of h0 to 120 F and 90 per cent relative humidity. The neutron detectors are designed for continuous operation in an environment of 175 F, 90 per cent ~1ative humidity, ana 60 psig. Engineered safety features equipment and vi: 1 instrumentation inside the reactor building are capable of performing required safety functions for conditions 40 psig, 26h F, and 100 per cent RH), which are in excess of the requirements of the loss-of-coolant accident (7.1.1 and 7.1.2). I Redundant instrument channels are provided for the reactor protection and safety features actuation syste=s. Loss of power to each individual reactor protection channel vill trip that individual channel. Loss of all instrument i j power vill trip the reactor protection system, thereby releasing the control l rods , and will activate the safety features actuation system instrumentation (with the exception of the containment spray valves). Manual reactor trip is designed so that failure of the automatic reactor trip circuitry will not prohibit or negate the manual trip. The same is true with respect to manual operation of the engineered safety features equipment. 1-29

D-B 1.k.27 CRITERION 27 - REDUNDANCY OF REACTIVITY CONTROL (Category A) At least two independent ranctivity control systems, preferable of different principles, shall be provided. Discussion This criterion is met by movable control rods and soluble boron poison (7 2.2.1). l.h.28 CRITERION 28 - REACTIVITY HOT SHUTDOWN CAPABILITY (Category A) At least two of the reactivity control systems provided shall be independently capable of making and holding the core suberitical from any hot standby or hot operating conditions, including those resulting from power changes sufficiently fast to prevent exceeding acceptable fuel damage limits. Discussion A single reactivity control system consisting of h9 full-length poison control rods is provided to rapidly make the core suberitical upon a trip signal. Trip signals are set to protect the core from damage due to the effects of any op-

erating transient. 'Ihe soluble absorber reactivity control system can add neg-ative reactivity to make the reactor suberitical. However, its action is slow, and its ability to protect the core from damage, which might result from rapid load changes such as a full load turbine trip, is not a design criterion for this system. The high degree of redundancy in the control rod system is con-sidered sufficient to meet the intent of this criterion (3.2.2.1). )

1.h.29 CRITERION 29 - REACTIVITY SEEDOWN CAPABILITY (Category A) At least one of the reactivity control systems provided shall be capable of making the core suberitical under any condition (including anticipated opera-tional transients), sufficiently fast to prevent exceeding acceptable fuel damage limits. Shutdown margins, greater than the maximum worth of the most effective control rod when fully withdrawn, shall be provided. Discussion The reactor design meets this criterion both under normal operating conditions and under the accident conditions set forth in Section lb. The reactor is de-signed with the capability of providing a shutdown ma gin of at least 1 per cent ak/k with the single most reactive control rod fully withdrawn at arr/ point in core life with the reactor at a hot, zero power condition. The mini-mum hot shutdown margin of 3.0 per cent Ak/k occurs at the end of life (3.2.2.1)

 .vith the single most reactive control rod fully withdrawn.

1.h.30 CRITERION 30 - REACTIVITY HOLDDOWN CAPABILITY (Category B) At least one of the reactivity control systems provided shall be capable of ma'k ing and holding the core suberitical under any conditions with appropriate margins.for contingencies. i v 1-30 , ,,

                                                          /f l

n-- - D-B Discussion The reactor meets this criterion with control rods for hot shutdown under nor-mal operating conditions and for shutdown under the accident conditions set forth in Section ik. Reactor suberitical margi'n is maintained during cooldown by changes in soluble boron concentration. The rate of reactivity compensation from boron addition is greater than the reactivity change associated with the reactor cooldown rate of 100 F/h. Thus, suberiticality .is assured during cooldown with the most reactive control rod totally unavailable (3.2.2.1). l.h.31 CRITERION 31 - REACTIVITY CONTROL SYSTEMS MALFUNCTION (Category B) The~ reactivity control systems shall be capable of sustaining any single mal-function, such as unplanned continuous withdrawal (not ejection) of a control rod without causing a reactivity transient which could result in exceeding ac-ceptable fuel damage limits. , Discussion The reactor design meets this criterion. A reactor trip will protect against continuous withdrawal of a control rod (1k.1.2.3). l.h.32 CRITERION 32 - MAXIMUM REACTIVITY WORTH OF CONTROL RODS (Category A) Limits, which include considerable margin, shall be placed on the maximum re-activity worth of control rods or elements, and on rates at which reactivity can be increased to insure that the potential effects of a sudden or large change of reactivity cannot: (a) rupture the reactor coolant pressure boundary or, (b) disrupt the core, its support structures , or_ other vessel intern'als sufficiently to impair the effectiveness of emergency core cooling. Discussion The reactor design meets this criterion by sa'fety features which limit the maxi-mum reactivity insertion rate. These include rod-group withdrawal interlocks, soluble . boron concentration reduction interlock, maximum rate of dilution water ' additon, and dilution-time cutoff (lk.l.2.h) . In addition,' the rod drives and , their controls have an inherent feature that limits overspeed in the event of malfunctions (3.2.h.3). Ejection of the maximum-worth control red will not lead to further coolant boundary rupture or to internals damage which would interfere with emergency core cooling (lh.2.2.2). l.h.33 CRITERION 33 - REACTOR COOLANT PRESSURE BOUNDARY CAPABILITY (Category A) The reactor coolant pressure boundary shall be capable of accommodating with-out rupture, and with only limited allowance for energy absorption through plastic deformation, the . static and dynamic loads imposed on any boundary com- l

ponent as a result of any inadvertent and sudden release of energy to the cool-ant. As a design reference, this sudden release shall be taken as that which.

would result from a sudden reactivity insertion such as rod ejection (unless prevented by positive mechanical means), rod dropout, or cold water addition. i

_ ~ ' 42

1-31'

D-B Discussion Tha reactor design meets this criterion. There are no credible mechanisms wh2reby damaging energy releases are liberated to the reactor coolant. Ej ec-tion of the maximum worth control rod will not lead to further coolant boundary rupture (1k.2.2.2). l.h.3h CRITERION 3h - REACTOR COOLANT PRESSURE BOUNDARY RAPID PROPAGATION FAILURE PREVENTION Th2 reactor coolant pressure boundary shall be designed to minimize the prob-ability of rapidly propagating type failures. Consideration shall be given (a) to the notch-toughness properties of materials extending to the upper shelf of the Charpy transition curve, (b) to the state of stress of materials under static and transient loadings, (c) to the quality control specified for materials and component fabrication to limit flav sizes, and (d) to the provisions for control over service temperature end irradiation effects which may require op-erstional restrictiens. Discussion The reactor coolant pressure boundary design meets this criterion by the fol-loving:

a. Development of reactor vessel plate material properties opposite the core to a specified Charpy-V-notch test result of 30 ft/lb or great-er at a nominal lov NDTT.

b. Determination of the fatigue usage factor resulting from expected static and transient loading during detailed design and stress analy-sis.

c. Quality control procedures including permanent identification of ma-terials and non-destructive testing.

  ,            d. Operating restrictions to prevent failure towards the end of design vessel life resulting from increase in the nil-ductility transition temperature (NDTT) due to neutron irradiation, as predicted by a
    ,             material irradiation surveillance program (k.h.3).

1.h.35 CRITERION 35 - REACTOR COOLANT PRESSURE BOUNDARY BRITTLE FRACTURE PREVENTION (Category A) j Under conditions where reactor coolant pressure boundary system components con-structed of ferritie materials may be subjected to potential loadings, suen as n reactivity-induced loading, service temperatures shall be at least 120 F l cbove the nil-ductility transition (NDT) temperature of the component material I if the resulting energy release is expected to be absorbed by plastic deforma- l tion, or 60 F above the NDT temperature of the component material if the re-sulting energy release is expected to be absorbed within the elastic strain energy range. 1 .

.)'

i 43 ! 1-32 1

D-B Dis cussion T The reactor vessel is the only reactor coolant system component exposed to a significant level of neutron irradiation and is, therefore, the only component subject to material irradiation damage. Unit cperating procedures will limit the operating pressure to 20 per cent of the design pressure when the reactor coolant system temperature is below NDTT +60 F throughout unit life. Analysis has shown no potential reactivity-induced conditions which will result in energy release to the primary system in the range expected to be absorbed by plastic deformation (h.3.3). l.h.36 CRITERION 36 - REACTOR COOLANT PRESSURE BOUNDARY SURVEILLANCE (Category A) Reactor coolant pressure boundary components shall have provisions for inspec-tion, testing, and surveillance by appropriate means to assess the structural and leak-tight integrity of the boundary components during their service life-time. For the reactor vessel, a material surveillance program conforming with ASTM-E-185-66 shall be provided. - Discussion The reactor coolant pressure boundary components meet this criterion. Space is provided for non-destructive testing during plant shutdown. A reactor pres-sure vessel material surveillance program conforming to ASTM-E-185-66 nas been established (h.h.3) . 1.h.37 CRITERION 37 - ENGINEERED SAFETY FEATURES FOR DESIGN (Category A) Engineered safety features shall be provided in the facility to back up the safety provided by the core desigr., the reactor coolant pressure boundary, and their protection systems. As a minimum, such engineered safety features shall be designed to cope with any size reactor coolant pressure boundary break up to and including the circumferential rupture of any pipe in that boundary as-suming unobstructed discharge from both ends. Discussion The reactor design meets this criterion. The emergency core cooling systems can protect the reactor for any size leak up to and including the circumferen-tial rupture of the largest reactor coolant pipe (1h.2.2.3). 1.h.38 CRITERION 38 - RELIABILITY AND TESTABILITY OF ENGINEERED SAFETY FEATURES (Category A) All engineered safety features shall be designed to provide high functional reliability and ready testability. In determining the suitability of a facil- . ity for a proposed site, the degree of reliance upon and acceptance of the inherent and engineered safety afforded by the systems, including engineered safety features, will be influenced by the known and the demonstrated per-formance capability and reliability of the systems, and by the extent to which the operability of such systems can be tested and inspected where appropriate i during the life of the plant. 44 / .

D-B s Discussion All engineered safety features are designed so that a single failure of an active component in a system will not prevent operation of that system or re-duce its capacity below that required to_ maintain ,a ,saf,e condition.. Two in, dependent containment vessel cooling systems, each having full heat removal

 ,        capacity, are provided to prevent overpressurization.

8 l The high pressure injection, core flooding, and low pressure injection systems hav; sepsrate equipment and instrumentation strings to ensure availability of capacity. Some portions of the engineered safety features have both a normal and an emergency function, thereby providing nearly continuous demonstration of oper-ability. During normal operation, the standby and operating units will be rotated into service on a scheduled basis. Engineered safety features equipment piping that is not fully protected against LOCA missile damage utilizes dual lines to preclude loss of the protective function as a result of the secondary failure. Testing and inspection of the engineered safety features is further described in Section 6. 1.h.39 CRITERION 39 - EMERGENCY POWER FOR ENGINEERED SAFETY FEATURES (Category A) i Alternate power systems shall be provided and designed with adequate independency, redundancy, capacity, and testability to permit the function-6 lingrequiredoftheengineeredsafetyfeatures. As a minimum, the onsite power system and the offsite power system shall each, independently, provide this capacity assuming a failure of a single active component in each power system. Diucussion Alternate power systems will be provided and desi 6ned with adequate indepen-dency, redundancy, capacity, and testability to permit the functioning of the engineered safety features. Alternate emergency power systems vill consist of:

a. Two sources of offsite power from the 3h5 kV svitchyard. Either source vill be capable of supplying power for the engineered safe-ty features of the station. The 3h5 kV switchyard and system vill be so arranged that a failure of a single active component will not prohibit functioning of both offsite power sources.

b. Two diesel generators, either of which will be adequate to supply two power required for the engineered safety features. Failure of either generator will not prohibit the functioning of the minimun safety features required to maintain the station in a s safe condition. )

45 Amendment No. 8 1-3h

L .B ('s 1

c. Two de batteries , either of which will be adequate to supply the de control power required for the engineereu afety features.

Failure of a single active component in this system vill not impair control of the minimum engineered safety features required to maintain the station in a safe condition. 1.4.k0 CRITERION h0 - MISSILE PROTECTION (Category A) Protection for engineered safety features shall be provided against dynamic effects and missiles that might result from plant equipment failures. Discussion Protective valls and slabs, local missile shielding, or restraining devices are provided to protect the containment vessel (5.2.1.9) and engineered safety features (5 2.3) within the containment vessel against damage from missiles generated by equipment failures associated with a loss-of-coolant accident. The concrete enclosing the reactor coolant system serves as radiation shielding and as an effective barrier against internal missiles. Local missile barriers are provided for control rod drive assemblies. The contai" ment vessel interior structure is designed to sustain loads which could result from fLilure in major equipment and piping, such as jet thrusts, jet impingement, and local pressure transients (5.2.3). 1.h.hl CRITERION hl - ENGINEERED SAFETY FEATURES PERFORMANCE CAPABILITY (Category A) . Engineered safety features, such as emergency core cooling and containment heat removal systems, shall provide sufficient performance capability to accommodate partial loss of installed capacity and still fulfill the required safety func-tion. As a minimum, each. engineered safety feature shall provide this required safety function assuming a failure of a single active component. Discussion All engineered safety features are designed so that a single failure of an active component vill not prevent operation of that system or reduce the sys-tem capacity below that required to maintain a safe condition. Redundancy is provided in equipment and pipelines so that the failure of a single active component of any system vill not impair the required safety function of that system. 1.h.h2 CRITERION h2 - ENGINEERED SAFETY FEATURES COMPONENTS CAPABILITY (Category A) Engineered safety features shall be designed so that the capability of each - ccmponent' and system to perform its required function is not impaired by the effects of a loss-of-coolant accident. 46 1-35

D-B

Discussion

The engineered safety features design meets this criterion. A single-failure 8 l analysis heat removalof the emergency systems core cooling

                                   /6.2.4) demonstrates      systems that these      (6.1.3.1),

systems and_ containm have sufficient redundancy to perform their design functions. The core flooding lines contain check valves which operate to permit flow of emergency coolant from the tanks to the containment vessel. These valves are self-actuating and need no external signal or external supplied energy to make them operate. Accordingly, it is not considered credible that they would fail to operate when needed. The engineered safety features are designed to function in the unlikely event of a loss of coolant accident with no impairment of 6.ction due to the ef-fects of the accident. 1.h.h3 CRITERION h3 - ACCIDENT AGGRMATION PREVENTION (Category A) Engineered safety features shall be designed so that any action of the engi-neered safety features which might accentuate the adverse after-effects of the loss of normal cooling is avoided. Discussion The engineered safety features are designed to meet this criterion. The water injected to ensure core cooling is sufficiently borated to ensure core sub- ) criticality. Non-essential sources of water inside the reactor building are automatically isolated to prevent dilution of the borated coolant. Essential sources of post-accident cooling waters are monitored to detect leakage which may lead to dilution of boron content. An analysis has been made to demon-strate that the injection of cold water on the hot reactor coolant system sur-faces will not lead to further failure. The design of the equipment and its actuating system ensures that water injection vill occur in a sufficiently short time period to preclude significant metal-water reactions and consequent

energy release to the containment (1k.2.2.3).

1.h.hh CRITERION hk - EMERGENCY CORE COOLING SYSTDG CAPABILITY (Category A) At least two emergency core cooling systems, preferably of different design principles, each with a capability for accomplishing abundant emergency core cooling, shall be provided. Each emergency core cooling system and the core shall be designed to prevent fuel and clad damage that would interfere with the emergency core cooling function and to limit the clad metal-vater reaction to negligible amounts for all sizes of breaks in the reactor coolant pressure boundary, including the double-ended ruptr.re of the largest pipe. The per-formance of each emergency core cooling system shall be evaluated conserva-tively in each area of uncertainty. The systems shall not share active com-ponents and shall not share other features or components unless it can be demonstrated that: (a) the capability of the shared feature or component to , perform its required function can be readily ascertained during reactor op- - ! eration, (b) failure of the shared feature or component does not initiate a loss-of-coolant accident, and (c) capability of the shared. feature or com- -] Amendment No. 8 1-36

                                                                                      . - . - . ~

l D-B

   <- ponent to perform its required function is not impaired by the effects of a                 ,

loss-of-coolant accident and is not lost during the entire period this fune- ' tion is required following the accident. Discussion Emergency core cooling is provided by pumped injection and pressurized core flooding tanks (6.1.3). Pumped injection is subdivided in such a way that there are two separate and independent strings, each including both high pressure and lov. pressure coolant injection, and each capable of providing 100 per cent of the necessary core injection with the core flooding tanks. There is no shar-ing of active components between the two subsystems in the post-accident oper-ating mode. The core flooding tanks are passive components which are needed for only a short period of time after the accident, thereby assuring 100 per cent availability when needed. This equipment prevents clad melting for the entire spectrum of reactor coolant system failures ranging from the smallest leak to the complete severance of the largest reactor coolant pipe. 1.h.h5 CRITERION h5 - INSPECTION OF EMERGENCY CORE COOLING SYSTEMS (Category A) Design provisions shall be made to facilitate physical inspection of all crit-ical parts of the emergency core cooling systems italuding reactor vessel in-ternals and water injection nozzles. Discussion All critical parts of the emergency core cooling systems, including the containment vessel internals, can be inspected during station shutdown {h.h).

l.h.46 CRITERION h6 - TESTING OF EMERGENCY CORE COOLING SYSTEMS COMPONENTS (Category A) Design provisions shall be made so that active components of the emergency core cooling systems, such as pumps and valves, can be tested periodically for op-erability and required functional performance. Discussion The design of emergency core cooling systems and components has incorporated adequate test and operational features to permit periodic testing of active components to assure operability and functional capability. Core _ flooding tank functional performance vill be demonstrated only in pre-operational test-ing. 1.h.kT CRITERION h7 - TESTING OF EMERGENCY CORE COOLING SYSTEMS (Category A) A capability shall be provided to test periodically the delivery capability of the emergency core cooling systems at a location as close to the core as is practical.

                    ~           b0 l-37
                         ~

m

D-B Discussion w

g The high pressure and low pressure injection (decay heat removal) sys-tems are included as part of normal service systems. Consequently, the active components can be tested periodically for delivery capability. TLL core flooding system delivery capability will be demonstrated during startup testing. In addition, all valves will be periodically cycled to ensure oper-ability. Witn these previsions , the delivery ca,3bility of the emergency core cooling systems can be periodically demonstrated (6.1.h). 1.b.k8 CRITERION h8 - TESTING OF OPERATIONAL SEQUENCE OF EMERGENCY CORE COOLING SYSTEMS (Category A) A capability shall be provided to test, under conditions as close to design as-practical, the full operational sequence that would bring the emergency core cooling systems into action, including the transfer to alternate power sources. Discussion The operational sequence that would bring the emergency core cooling systems into action, including transfer to alternate power sources, can be tested in parts (6.1.h and 7.2.3). l.h.h9 CRITERION h9 - CONTAINMEh"I' DESIGN BASIS (Category A) The containment structure, including access openings and penetrations, and any necessary containment heat removal systems shall be designed so that the j containment structure can accommodate without exceeding the design leakage rate the pressures and temperatures resulting from the largest credible energy release following a loss of coolant accident, including a considerable margin for effects from metal-water or other chemical reactions that could occur as a consequence of failure of emergency core cooling systems. Discussion The containment structure, including access openings and penetrations, is designed to accommodate, without exceeding the design leak rate, the transient peak pressure and temperature associated with a postulated piping break up to and including a double-ended rupture of the largest reactor coolant pipe. The containment structure and engineered safety features systems will be evaluated for various combinations of ener y release. The analysis will account for system energy and decay heat. The emergency injection system is

   -designed such that no single failure could result in significant metal-water reaction. The cooling capacity of either the containmer.t cooling system or the containment spray system (see Criterion 52) is adequate to prevent over pressurization of the structure, and to return the containment to near atmospheric pressure. The details of this evaluation are discussed in Section lb. Safety features design is discussed in Section 6.

Electric motors and valves which must fonction within the containment during accident donditions, will be capable of operating in the steam-air atmosphere , at the containment design pressure and temperature, j Amendment No. 8 1-38 4g

D-B 1.h.50 CRITERION 50 - NDT REQUIREMENT FOR CONTAINMENT MATERIAL [' (Category A) Principal lead carrying components of ferritic materials exposed to the external environment shall be selected so that their temperatures under normal opeating and testing conditier.s are not less than 30 F above Nil Ductility Transition (NDT) temperature. Discussion The containment vessel is not exposed to the external environment but to the environment of the annulus between the vessel and the shield building. All ferritic materials entering into the fabrication of the containment will have a NDT temperature at least 30 F below the temperature for operating and test-ing. l.h.51 CRITERION 51 - REACTOR COOLANT PRESSURE BOUNDARY OUTSIDE CONTAIUMENT (Category A) If part of the reactor coolant pressure boundary is outside the containment, appropriate features, as necessary, shall be provided to protect the health and safety of the public in case.of an accidental rupture in that part. Deter-mination of the appropriateness of features , such as isolation valves and ad-ditional containment, shall include consideration of the environmental and population conditions surrounding the site. Discussion The reactor design meets this criterion. The reactor coolant pressure boundary is defined as those piping syste=s or co=ponents which contain reactor coolant at design pressure and temperature. With the exception of the reactor coolant sampling line, the reactor coolant pressure boundary, as defined above, is lo-cated entirely within the Containment Vessel. The sampling line is provided with remotely operated valves for isolation in the event of a failure. This line is normally isolated and is used only during actual sampling operations. All other piping and components which may contain reactor coolant are at lov temperatures such that any leakage vor'd be collected by the contaminated drain system. No significant environmental 'ese would arise from these sources (9 7). 1.4.52 CRITERION 52 - CONTAINMENT HFAT REMOVAL SYSTEMS (Category A) Where active heat removal systems are needed under accident conditions to prevent exceeding containment design pressure, at least two systems, preferably of different principles, each with full capacity, shall be provided. Discussion The containment spray system, consisting of two pumps and two spray headers and the containment air recirculation system, consisting of two of three fan coolers, function as emergency heat removal systems. Each of t'ne two systems has the full heat removal capability required for the loss-of-coolant accident conditions (6.2). 50 1-39

g -

1 u ' d D-B

                                                                                              . -)'

1.k.53 CRITERION 53 - CONTAINMENT ISOLATION VALVES (Category A) l Penetrations that require closure for the containment function shall be l protected by redundant valving and associated apparatus. Discussion The general design basis governing isolation is that leakage through all fluid l penetrations not serving accident-consequence-limiting systems is to be mini-mized by a double barrier so that no single, credible failure or malfunction of an active component can result in loss-of-isolation or intolerable leakage. The installed double barriers take the form of closed piping systems, both inside and outside the containment. vessel, and various types of isolation valves. l.h.5h CRITERION 5h - CONTAINMENT LEAKAGE RATE TESTING (Category A) Containment shall be designed so that an integrated leakage rate testing can be conducted at design pressure after completion and installation of all l penetrations and the leakage rate measured over a sufficient period of time to verify its conformance with required performance. Discussion ! The containment vessel is designed so that initial integrated leak rate i testing at design pressure can be performed after completion and installation 6 lof penetrations and equipment (5.8.1.2,15.h.h). 3 1.h.55 CRITERION 55 - CONTAINMENT PERIODIC LEAKAGE RATE TESTING l (Category A) The containment shall be designed so that integrated leakage rate testing can be done periodically at design pressure during plant lifetime. l ! Discussion

The containment will be designed to permit periodic leak rate testing at the design pressure. However, leak rate testing vill be done at reduced pressure.

The acceptable leak rate at the reduced pressure vill be determined from results of the initial leak rate . test. Frequency of the tests vill be i dependent upon available margin between test values of leakage and the . 6 l acceptable limit- (5.8.1.2,15.h.h) . 1.k.56 CRITERION 56 - PROVISIONS FOR TESTING OF PENETRATIONS (Category A)

         - Provisions shall be made for testing penetrations which have resilient seals or-expansion bellows to permit leak tightness to be demonstrated at design pressure at any time.

Discussion Penetrations having resilient seals or expansion bellows may be tested per--

          'iodically at the design pressure (5.8.1.1.2).                                        ]
         . Amendment No. 6                        1-h0              51

D-B s 1.h.57 . CRITERION 57 - PROVISION 3 FOR TESTING OF ISOLATION VALVES (Category A) Capability shall be provided for testing functional operability of valves and a3sociated apparatus essential to the containment function for establish-ing that no failure has occurred and for determining that valve leakage does not exceed acceptable limits. Discussion Each containment isolation valve vill be tested periodically during normal operation or during shutdown conditions to ensure its operability and to ensure that valve leakage does not exceed acceptable limits (5.3). 1.4.58 CRITERION 58 - INSPECTION OF CONTAINMENT PRESSURE-REDUCING SYSTEMS (Category A) Design provisions shall be made to facilitate the periodic physical inspection , of all important components of the containment pressure-reducing systems, such as , pumps , valves, spray nozzles, torus , and sumps. Discussion The containment vessel spray system (6.2.2) essential equipment, except risers, distribution header piping, spray nozzles and the containment sump are located outside of the containment. The containment vessel sump and the spray piping and nozzles can be inspected during shutdown. Associated equipment outside the shield building can be visually inspected at any time. The containment air reciruelation system (6.2.3) has coolers and blowers inside the containment vessel and they can be inspected periodically. The service water system is outside of the containe nt and can also be inspected. l.h.59 CRITERION 59 - TESTING OF CONTAINMENT PRESSURE-REDUCTING SYSTEMS COMPONENTS (Category A) The containment pressure-reducing systems shall be designed so that active coinponents, such as pumps and valves, can be tested periodically for operability and required functional performance. Discussion The containment vessel air recirculation and cooling system is normally in service. Valving on the coils can be periodically cycled, thus placing the coils into emergency service, periodically during operation. The active com-ponents of the containment vessel spray system are tested periodically as set forth in 6.2.2.3. 1.k.60 CRITERION 60 - TESTING OF CONTAINMENT SPRAY SYSTEMS (Category A) A capability shall be provided to test periodically the delivery capability of the containment spray system as a position as close to the spray nozzle

 ,   as-is practical.

52 l bl o

D-B Discussion The delivery capability of the containmen't spray system vill be tested periodically to the extent practical up to the isolation valves before the spray nozzles. That portion of the system in' side the containment will be tested during shutdown for flow and distribution using air and air flow telltales (6.2). l.h.61 CRITERION 61 - TESTING OF OPERATION SEQUENCE OF CONTAINMENT PRESSURE-REDUCING SYSTEMS (Category A) A capability shall be provided to test, under conditions as close to the design as practical, the full operational sequence that would bring the containment pressure-reducing systems into action, including the transfer to alternate power sources. Discussion The containment cooling units (6.2) which are normally in operation can be switched into the accident mode of operation at any time. The standby unit can be started and tested at any time. Transfer to alternate power sources can also be tested. The operational sequence that would bring the contain-ment spray system into action, including the transfer to alternate power sources, can be tested in parts. 1.h.62 CRITERION 62 - INSPECTION OF AIR CLEANUP SYSTEMS (Category A)

Design provisions shall be made to facilitate physical inspection of all 8 lcriticalpartsofcontainmentaircleanupsystems,suchasducts, filters, fans, and dampers. Discussion There 1s no loss-of-coolant accident air cleanup system installed within the containment vessel. A shield building and penetration room ventilation and filtration system (6.3) , are provided for control of contaminants which might leak from the containment vessel following a loss-of-coolant accident. All components of the shield building and penetration room ventilation and filtration system including ducts, filters, fans, and dampers ~are located out-side the shield building filtration region. Access for physical inspection of these components is thereby not impeded by reactor operation. 1.h.63 CRITERION 63 - TESTING OF AIR CLEANUP SYSTEMS COMPONENTS (Category A) Design provisions shall be made so that active components of the air cleanup < systems, such as fans and dampers, can be tested periodically for operability and required functional performance. Amendment No. 8 53 1-k2

D-B Discussion Active components of the shield building and penetration room ventilation and filtration system (6.3) can be tested periodically for operability and required functional performance. 1.h.6h CRITERION 64 - TESTING OF AIR CLEANUP SYSTEMS (Category A) A capability shall be provided for in-site periodic testing and surveillance of the air cleanup systems to ensure (a) filter bypass paths have not developed and (b) filter and trapping materials have not deteriorated beyond acceptable limits. Discussion A capability for periodic testing and surveillance of the shield building and penetration room ventilation and filtration system (6.3) to ensure that (a) - filter bypass paths have not developed and (b) filter and trapping materials have not deteriorated beyond acceptable limits is provided. Filters in the system vill be tested and inspected periodically. These are outside the shield building and available for t?". ting and inspection at any time, l.h.65 CRITERION 65 - TESTING OF OPERATIONAL SEQUENCE OF AIR CLEANUP SYSTEMS (Category A) A capability shall be provided to test under conditions as close to design as practical and full operational sequence that would bring the air cleanup systems into action, including the transfer of alternate power sources and the design air flow delivery capability. Discussion The full operational sequence that would bring the shield building and pene-tration room ventilation and filtration system (6.3) into action, including the transfer to alternate power sources and the design air flow delivery capability, can be tested, l.h.66 CRITERION 66 - PREVENTION OF FUEL STORAGE CRITICALITY (Category B) Criticality in new and spent fuel storage shall be prevented by physical systems or processes. Such means as geometrically safe configurations shall be. emphasized over procedural controls. Discussion " This criterion is met by the design of the new or spent fuel assembly storage facilities (9.6) to maintain a safe condition by storing fuel assemblies in racks having spacing and/or poison sufficient to maintain a k,ff of less than 0.90 when vet, s. 54 1-k3 m

D-B 1.h.67 CRITERION 67 - FUEL AND WASTE STORAGE DECAY HEAT (Category B) / Reliable decay heat removal systema shall be designed to prevent damage to the fuel in storage facilities that could result in radioactivity release to plant operating areas or the public environs. Discussion The decay' heat removal systems vill be designed to prevent damage to the fuel in storage facilities and prevent undue risks to the station operating areas or the public environs (9.k). 1.k.68 CRITERION 68 - FUEL AND WASTE STORAGE RADIATION SHIELDING (Category B) Shielding for radiation protection shall be provided in the design of spent fuel and vaste storage facilities as required to meet the requirements of 10 CFR 20. Discussion The shielding provided in the spent fuel and vaste storage area is in accordance with the radiation zoning described in Section 9.6, enabling the station to meet the guidelines of 10 CFR 20. 1.k.69 CRITERION 69 - PROTECTION AGAINST RADI0 ACTIVITY RELEASE FROM I SPENT FUEL AND WASTE STORAGE (Category B) Containment of fuel and vaste storage shall be provided if accidents could lead to release of undue amounts of radioactivity to the pub!.ic environs. Discussion The speat fuel storage pool (9.6) is located within the fuel handling and storage area of the Auxiliary Building. The liquid vaste processing equip-ment and the gaseous vaste storage and disposal equipment is located within separate area of the same building (11.1). Both of these areas provide con-fi.nement capability in the event of an accidental release of radioactive materials, and both are ventilated with discharges which are monitored (9.8). Analysis has indicated that the accidental release of the maximum activity content of the vaste gas decay tanks will not result in excessive dosec (1ks2.2 5).

Radioactive liquid erfluent leakage into the component cooling water system vill be determined by radiation monitors. Any accidental leakage from liquid vaste processing equipment vill be collected in e sump and transferred to tanks to prevent releases to the environment.

he _. 55 1-kh

           .esasa        e+                        .

D-B l.h.70 CRITERION TO - CONTROL OF RELEASES OF RADIOACTIVITY TO THE ENVIRONMEN'I' (Category B) The facility design shall include those means necessary to maintain control over the plant radioactive effluents, whether gaseous, liquid, or solid. Appropriate holdup capacity shall be provided for retention of gaseous, liquid, or solid effluents, particularly where unfan rable environmental conditions can. be expected to require operational limitations upon the release of redioactive effluents to the environment. In all cases, the design for radioactivity control shall be justified (a) on the basis of 10 CFR 20 requirements for normal operations and for any transient situation that might reasonably be anticipated to occur, and (b) on the basis of 10 CFR 100 dosage level guidelines for potential reactor accidents of exceedingly low probability of occurrence except that reduction of the recommended dosage levels may be required where high population densities or very large cities can be affected by the radioactive effluents. Discussion The radioactive vaste system vill collect, segregate, process, and dispose of radioactive solids , liquids , and dases in such a manner as to comply with 10 CFR 20 (11.1). Solid vastes will be processed in a batch manner for off-site disposal. Pro-cessed liquid vastes and gaseous vastes released to the environment will be monitored and discharged with suitable dilution to assure tolerable activity levels on the site and at the site boundary. Ample holdup storage capacity for liquid and gaseous vaste is provided. Wastes are sampled to establish release rates consistent with environmental conditions. In-line monitoring vill provide a continuous check on the release of activity. Under accident conditions, radioactive gaseous effluents which may be released into enclosed areas are collected by the ventilation systems and discharged to the station vent (9.8). Permanently installed area detectors and the station vent detectors are used to monitor the discharge levels to the environment. In addition, portable monitors are available on site for supplemental surveys. 56

r D-B 15 RESEARCH AND DEVELOPMENT REQUIREMENTS The research and development program; that have been initiated to establish final design or to demonstrate the capability of the design for future opera-l tion at a higher power level are summarized as 0,llows: L 1 5.1 XENON OSCILLATIONS Spatial xenon instability is not a safety problem in itself, for if an oscil-lation occurred and were allowed to continue, the plant would eventually be tripped by the high flux scram before unacceptable fuel damage could occur. The period of oscillation of xenon induced spatial instabilities is long, 25 to 30 hours. Therefore, it can be thwarted or controlled by operator action using pre-determined rod manipulations. In the extreme condition, the plar.t can be shutdown and no safety problem would result. Xenon stability is an operational consideration and the stability character-istics of the core are analyzed in regard to steady state and transient oper-ation. These effects must be known in order to ;atrize both power production and plant availability by allowing the judiciou. action of a maneuvering and control system and the inclusion of modifications if necessary to ensure the desired core characteristics. This progrts is concerned with establishing the stability of the core and with , evaluating the effects of part length control rods and burnable poison clusters ; on core stability. If mechanisms for control of diverging xenon oscillations are required they will be developed. T;ie control study vill also define a method for detection of oscillatory behavior. The xenon program consists of the following:

a. Modal analysis.

b. One- and two-dimensional digital analysis.

c. Three-dimensional digital analysis.

The results of the modal analysis have been submitted as Topical Report BAW-10010 " Stability Margin for Xenon Oscillations -- Modal Analysis." One-dimen-sional digital analysis is to be used to ascertain the validity of the modal analysis approach. Two-dimensional analysis is to be performed in both RZ and XY geometries with feedback mechanisms--in RZ to look at the relation of the 1-D axial stability to finite dimensions and in XY to determine the azimutha.1 characteristics of the core which can then be compared to the results of the three-dimensional analysis. Thus, the objective of this phase of the program is to relate the digital analysis to the modal analysis. The reference core vill be analyzed 'in a three-dimensionnl digital analysis with nuclear-thermal feedback. The results of the one- and two-dimensional digital analyses vill be filed as , a topical report in the third quarter of 1969 and the'results of the three- ,)l dimensional. digital analyses vill be filed when completed for Davis-Besse. 57 1 k6

D-B (~ The program is scheduled for completion well before any constraining date in regard to plant construction and final core design specifications. 1.5.2 THERMAL AND HYDRAULIC PROGRAMS B&W is conducting a continuous research and development program for heat trans-fer and fluid flow investigations applicable to the design of the Davis-Besse station. The thermal and hydraulic design limits, which establish the experi-mental program criteria, are discussed in paragraphs 3.1.2.3 and 3.; 1.1. The only required information for licensing at the design core power level, using the W-3 correlation, is the reactor vessel model flow test information to substantiate the core thermal and hydraulic design. A 1/6-scale model of the vessel and internals was under test to measure the flow distribution'to the core, fluid mixing in the vessel and core, and the distribution of pressure drop within the reactor vessel. . All of the tests, relating to the safety analysis, maximum design power rating and fabrication of reactor internals, have been satisfactorily completed. Test data analysis and documentation were conduc',ed and a final report was submitted as Topical Report BAW-10012, " Reactor Vessel Model Flow Test." The R&D programs will be judged adequate if, utilizing the W-3 correlation and the analysis of vessel model flow test data, it is confirmed that the core heat transfer design is safely belov Departure from Nucleate Boiling (DNB) at the design overpower (112 per cent of rated power). See paragraphs 3.1.2.3 and 3.2.3.2.3. Since conservative assumptions have been made throughout the de-sign, and test work appears to confirm it, there is no reason to expect that the R&D will prove inadequate. 1.5.3 FUEL R0D CLAD FAILURE" A study of clad failure mechanisms associated with a loss-of-coolant accident is presently under way. This study has included identification of the poten-tial failure mechanisms, a search of the literature to obtain applicable data, evaluation and application of existing data, and scoping tests to obtain data on potentisl failure mechanisms. The initial results of this study include the identification of the failure mechanisms, an evaluation of the information available in the literature concerning these mechanisms, and an evaluation of the effects of these mechanisms en the reactor system design. The objective of the study is to insure that there are no potential failure mechanisms that might interfere with the ability of the emergency core cooling system to terminate the core temperature transient and remove decay heat in the eventlof a loss-of-coolant accident. These potential failure mechanisms include clad melting, :irconium-water reaction, eutectic formation between the 4 Zircaloy-clad and the stainless steel spacer grids, the possibility of clad embrittlement as a result of the quenching during core flooding, and clad per-foration or deformation accompanying its failure. In t:e case of clad melting ACRS asterisked item. 58 l 1 h7 e .

D-B and zirconium-water reaction, our present design limit for peak clad tempera- j ture precludes these as possible failure modes. Information available in the literature, along with experimental evidence from tests conducted by B&W, show that brittle fracture of the cladding vill not occur as a result of quenching following a loss-of-coolant accident, and that eutectic formation between dis-l similar core materials will not interfere with the flow of emergency core cool-ant after the accident. Tne nature of tests is discussed in paragraph 1h.2.2.3.h and. the need for the tests is discussed in the following paragraphs. The tests are intended to demonstrate that in the event of a LOCA, the amount and locationsxt expansion of the cladding due to a net internal pressure vill not cause a blockage of coolant channels which would permit a peak clad temperature in excess of that permitted by the ECCS deeign criteria. The criteria for emergency core cool- ! ing system performance, including the clad temperature limit, are discussed in paragraph 6.1.1. Preliminary tests have shown that clad expansion is localized, and that any sig-nificant pressure is relieved by perforation at temperatures of the order of 1,000 to 1,400 F. The force for continued expansion therefore is dissipated. Extensions of these preliminary tests are scheduled to evaluate the effects of other variables in order to verify the conclusion that coolant channels will remain sufficiently open to permit core cooling. Phase I of the detailed experimental program vill consist of single pin tem-perature excursions, to supplement the preliminary clad swelling tests and will T investigate the effect of the various operating parameters (i.e. , temperature, / internal gas' pressure, clad hydrogen content, and degree of oxidation) on the clad failure mechanism. Foreover, efforts will be made to correlate failure modes with tubing dimensional characteristics as well as metallurgical or chem-ical properties. Phase II will consist of multirod tests to study the effect of the restraining action of adjacent fuel pins on the pin deformation as well as to confirm that deformation location vill be random so as not to be detrimental to the emer-gency cooling of the core during a loss-of-coolant accident (LOCA). An analytical study applying the results of Phase I and II to the loss-of-cool-ant accident analysis vill comprise Phase III. A mathematical model is being

   ,    developed to evaluate the effect of clad swelling on the fuel and clad temper-ature during a LOCA.

The test program is designed to provide the following information:

a. Clad burst temperature as a function of initial internal pressure.

b. Clad burst temperature as a function of burst pressure.

c. Clad deformation as a function of burst te=perature and initial pressure,

d. Effect of heatup rate on burst temperature and deformation.

e. Effect of. hydrogen content on burst temperature and deformation.

1 h8 -

                    -~                                                                 2- - x D-B

f. Effect of oxidation condition on the burst temperature and deforma-tion,

g. Correlation of deformation locations,

h. Restraining action of adjacent pins on deformation of deforming pin (h x 4 arrays).

1. Flow channel-blockage as well as mode and distribution of failed rods for worst conditions determined from Phase I of the tests.

J. A determination of effects of clad expansion modes on the fuel and clad temperature as a function of time during a LOCA. The program was designed on the basis that the major unknown was the amount and location of blockage that could result from the clad deformation in a LCCA. The parameters included in the single pin experimental test are all of those of primary importance with respect to the extent of clad perforations and the mode of clad failure. The multi-pin test will provide the data to de-termine the interaction between pins undergoing a temperature excursion and vill

describe the blockage mechanism. With these data available, an analysis of the capability of the ECCS to function as designed vill be verified. Data presently available indicate that the cladding deformation vill be of a random nature and of small =agnitude. The interpretation of these data leads to the conclusion that this phenomenon vill not affect ECCS performance sig-nificantly. Thus, the above outlined testing is of a confirmatory nature to more specifically evaluate the effect of clad swelling on the fuel and clad temperature during a LOCA. In view of the favorable results of the tests already performed by B&W, and similar results from independent tests,* the completion of B&W's current pro-gram is expected to provide further confirmation that coolant channel restric-tions due to clad swelling vill not limit ECCS effectiveness. Completion of the present program is expected in late 1969 which is sufficiently in advance of the scheduled Davis-Besse startup. 1.5.h. CONTROL ROD DRIVE LINE TEST The test assembly for this program was a full-sized fuel assembly vith asso-c!'_eed control rod and control rod guide, adjacent internals, and control rod drive. The purpose of this progrm vas to seek out potential material and/or design problems prior to production unit testing. The operational and life testing of the roller nut control rod drive mechanism was conducted in order to verify that: APED'5479, " Fuel Rod Failures During Si=ulated Loss-of-Coolant Conditions." l h9 '

D-B N,

a. Tht trip time of the mechanism meets the design requirements for a safe and orderly. shutdown on a complete loss of power to the plant, i.e., the-loss of the four primary coolant pumps.

b. The materials utilized in the construction of the mechanism are com-patible with the primary coolant chemistry to ensure the structural adequacy of the mechanism.

c. The wear characteristics of the component parts of the mechanism are satisfactory for the operational requirements.

   ; The drive test program was conducted in an autoclave under reactor operating conditions of temperature, pressure, flow and water chemistry. In addition, the guidance system was established with components which duplicate the guid-ance in the reactor through the full stroke of the mechanism; i.e., with a prototype fuel assembly and with a prototype upper control rod guide assembly.

The following information, which was obtained from the operational and life tests of the mechanism, has been considered in the safety evaluation of the mechanism:

a. The trip time of the mechanism associated with the 100 per cent max-imum internals misalignment conditions of the drive line.

b. The degree of wear exhibited by components of the mechanisms after )

life testing. The test program of the roller nut mechanism was performed by the B&W Research Center.in Alliance, Ohio in sufficient scope and depth to establish that the performance of the mechanism is satisfactory. Topical Report BAW-10007, " Con-trol Rod Drive System Test Program," provides the results and analysis'of the test data.

 ,   1.5 5        ONCE-THROUGH STEAM GENERATOR TEST Testing necessary to prove the adequacy of the once-through steam generator design for service at the initial power level and to confirm the size and con-figuration of_the units has been completed.
   . The areas of concern and investigation which made up this program are as fol-lows:

a. The steady state and transient operation tests have confirmed the analytically predicted performance characterisites of the steam generator and have provided a means for developing a satisfactory control scheme.

l

b. 'Feedvater spray nozzle tests have demonstrated that the design ar-rangement will heat the feedvater in a satisfactory manner.

I

c. Tube leak simulation tests have demonstrated that a leak in one tube

                                                                                      ~'
                                                                                         )

will not be propagated by causing a failure'in adjacent tub'es. 1-50 .

D-B { ' t d. Mechanical tests have demonstrated that the tubes can with:tand, without failure, the mechanical loads they may experience either during normal operation or accident condition, including primary side blowdown.

e. Vibration testing has demonstrated that the unit contains no un-desirable resonance characteristics.

f. Tests have been conducted to simulate a steam line failure or a re-actor coolant system failure to demonstrate that the unit will re-main intact under-conditions of rapid depressurization and large tem-perature differentials between the tubes and the shell.

The following general criteria guided the overall R&D program for the steam generator which has been completed:

a. To confirm analytical heat transfer and hydraulic performance char-acteristics of a steam generator utlizing the once through principle to provide slightly superheated steam while using a pressurized water reactor as the heat source.

b. To provide sufficient data to permit development of a control system which is compatible with the once through steam generator concept.

c. To investigate the various safety matters pertinent to the once

                  ' through steam generator design having a straight tube configuration and to demonstrate that no significant safety problems exist due to the unique design of this generator.

The results of the tests have been evaluated to the extent necessary to estab-lish that the above criteria have been met and to establish final design char-acteristics for manufacture of the steam generators. Thus, the design of the steam generators is based on data already available. Topical Report BAW-10002, "Research and Development Report for the Once-Through Steam Generator," presents the results of the once-through steam generator R&D. 1.5.6' SELF-POWERED DETECTOR TESTS . The results of the self-powered detector test program-are provided in Topical Report BAW-10001, "Incore Instrumentation Test-Program." The self-powered detector tests censisted of qualification testing with sur-ficient longevity to ensure that both neutron flux and power information can be reliably-measured. The self-powered detectors now en test have received an_ integrated dose which is equivalent to over four full power years at Davis-Besse conditions.and have shown no fault that would prohibit their use in PWR service. The self-powered detectors are not employed in any direct safety actions. Knowledge derived from these detectors will be used to provide flux and power distribution data for the. operator and the fuel management program and to help confirm physics calculations. I hL _

D-B The acceptance criteria for the self-powered detectors are to measure neutron flux with a relative accuracy of 5 per cent in a PWR environment over a three- '} year time span. The qualification testing of the self-powered detectors has indicated that this device is capable of measurin6 neutron flux in a PWR as well as, if not better, than any of the alternate devices previously considered. Its only limitation is its inherent slow time constant. However, for its service of measuring flux for operator use, fuel management calculations and physics data, the slow re-sponse is not a serious limitation. 1.5.7 BLOWDOWN FORCES ON INTERNALS AND CORE" The loads on the reactor and internals following a LOCA and the resultant stresses and deflections in the ree.ctor internals have been analyzed for B&W Huclear Supply Systems with skirt supported vessels and the results reported in Topical Report BAW-10008, " Reactor Internals Stress and Deflection Due to a Loss-of-Coolant Accident'(LOCA) and Maximum. Hypothetical Earthquake." An investigation has also been performed to determine the stresses and deflections in the core of skirt supported vessels and vill be submitted as a Topical Re-port in late 1969 Similar reports will be submitted for the Lavis-Besse noz-zle supported vessel as the analysis is completed. s ACRS asterisked item. - 1-52 b

D-B l.6 PROPOSED STATION DESIGN IN THE AREAS OF CONCERN IDENTIFIED IN ACRS LETTERS AS ASTERISKED ITEMS Item 1 - Fuel Clad Failure During a LOCA and Its Effect 'on ECCS Performance - The ACRS has requested further evidence, both analytical and experimental, that bulging and/or perforation of the fuel cladding following a IDCA vill not af-feet the heat transfer capability of ECCS to the extent that clad melting can occur. Response - The analytical and experimental work being carried out by B&W is described in Section 1.5.3 Item 2 - Partial Melting of Fuel Assembly During Normal Operation - The ACRS desires further information that the melting and subsequent disintegration of a portion of a fuel by inlet coolant orifice. blockage or by other means vill not lead to unacceptable conditions in terms of fission product release, local . high pressure production, and possible initiation of failure in adjacent fuel elements. Response - B&W has completed an analysis of sustained DNB operation caused by abnormal flow conditions without nuclear feedback for a core power of 2,568 MWt. Based on the results of the investigation the following conclusions can be reached:

a. A DNB which occurs on one fuel rod due to a flow blockage vill not propagate to adjacent fuel rods.

b. The maximum cladding temperature which would occur on a fuel rod in sustained DNB due to a flow blockage is 1,111 F.

c. The maximum fuel temperature which would occur in a fuel rod in sus-tained DNB due to a flow blockage is well below the melting point for UO2 -

d. A corrosion reaction sufficiently rapid to cause a sudden energy re-lease would not occur and the modh of cladding failure due to corro-sion-erosion would probably be a slov local failure.

e. Short-term cladding strength is sufficient to prevent cladding burst due to the internal pressure which would occur.

This analysis has been described in detail in the Topical Report BAW-100lk,

      " Analysis of Sustained Departure From Nucleate Boiling Operation."

Item 3 - Ability of Fuel to Withstand Expected EOL Transients - The ACRS recom-mends fuel burnup tests at linear heat generation rates higher than those cal-culated for the worst anticipated transient and. to burnups comparable to the maximum expected. l Response - In normal steady-state operation linear heat generation rates in Davis-Besse reach 17.8 kW/ft at the hottest spot in the core at initial rated

    . power. During vorst anticipated transients due to equipment malfunction or operator error, heat generation rates can reach the over-power condition of 19.9 kW/ft.                                                                                                 '

l 64 ! l-53 L

D-B It is B&W's judgment that sufficient data exist to justify operation at the ~'\ design power density for Davis-Besse. B&W is conducting a program to obtain / additional information on fuel growth rates and irradiation effects on clad-ding, the influence of hydrogen on cladding, and fission gas release at high burnup. The purpose of this program is to provide data for future operatir.g conditions. However, it vill also provide additional confirmation of operating conditions currently being licensed. Item h - Prima' ry System Quality Assurance and In-Service Inspection - The ACES vants every precaution possible taken to assure that a high quality primary sys-tem is fabricated and they further desire that critical areas of the primary eystem be inspectable during service life. Response - The overall quality assurance program is described in Appendix 1B. Quality control for the reactor coolant system is described in Section k.5. The in-service inspection is Gescribed in Section h.h.l. Item 5 - Effects of Thermal Shock on Pressure Vessel Integrity - The ACRS is concerned that the thermal shock resulting from actuation of the ECCS follow-ing a LOCA vill cause a flaw in the vessel to propagate through the vessel thickness. If this happened, the coolant would be lost and the core would be uncovered. Response - The analytical work has been completed and is presented in Topical Report BAW-10018, " Analysis of Structural Integrity of the Reactor Vessel Subjected to Thermal Shock." Stress intensity factors were determined by elas-ticity solutions for the particular geometry and loading conditions of the ) structure. The two methods employed in this analysis are based on fracture mechanics analyses using the Paris and Irwin equations. Both solutions apply super-position for a wedge-type load on a crack surface with corrections for two free surfaces which are present in the finite width vessel vall. The two solutions have been based on the stress distribution being caused by a mechani-cal load. Thus, no consideration has been made for the relieving of thermal stress due to the presence of the crack. This adds to the conservatism of the solution. The fracture toughness values used in the analysis are dependent upon both the temperature and integrated irradiation dosage. Since a limited amount of data are availat '.e, fracture toughness values at elevated temperatures were found by using valid test data at the maximum temperature at which it was obtained and conservatively extrapolating these data in a linear manner with tempera-ture. Item 6 - Effects of Blowdown Forces on Core and Internals - The ACRS desires that calculational models be developed and used to analyze in detail the ef-fects on core and internals of the blowdown forces experienced by those com-ponents during a LOCA. Response - The B&W program on effects of blevdown forces on the core and in-ternals is presented in Section 1.5.7 Item 7 - Separation of Control and Protection System Instrumentation - The Com-3 mittee believes that control and protection instrumentation should be sep- is / arated to the fullest extent practicable. Interconnection between control and 1-5h ma _4

-- - - -- . u=~ ~ D-B protection instrumentation should be eliminated or reduced to a minimum. The

applicant should review the design for common failure modes, taking into ac-count the possibility of systematic, non-random, concurrent failures of re-dundant devices, not considered in the single failure criterion. Resconse - The B&W system design meets the requirements for separation of pro-tection and control as specified in AEC General Design Criterion 22, the IEEE-279 Proposed Criteria for Nuclear Power Plant Protection Systems, and the requirements as stated in the United States position of July 1967, with respect to Section 5.k.1.1 of the International Electro-Technical Commission Document b5A. In addition, the system meets the Single Failure requirements of AEC General Design Criterion 21 and of the IZEE-279 proposed criteria. The pro-tection and control system designs are being analyzed for possible " common failure modes." Item 8 - Instrumentation for Prompt Detection of Gross Fuel Failure - The ACRS desires that consideration be given to the development of instrumentation to detect gross fuel element failure. This instrumentation should be capable of rapidly detecting fuel failure in the presence of fission products already in the coolant due tc " leakage" through the clad and other normally expected sources and to scram on the signal. Response - B&W has undertaken a scoping study on the problem of prompt detec-tion of gross fuel element failure. This study will describe methods available for detection of fuel failure by measuring reactor coolant activity levels. This study will use available data and technology to obtain a comparison of the postulated coolant activity due to fuel failure with that due to expected and potential sources of background activity. The study report is being pre-pared and will be issued soon. s e e 66 1-55

D-B s 17 THE TOLEDO EDISON COMPANY COMPETENCE TO il BUILD AND OPERATE DAVIS-BESSE NUCLEAR

            , POWER STATION Toledo Edison has many years experience in the design, construction and

, operation of fossil fuel fired generating unite and has a good reservoir of experienced and competent personnel in all phas2s of generating station and utility sys. tem design and operation. The Company has for a number of years participated in the Enrico Femi Fast Breeder project and nas key personnel with considerable experience in all phases of that project. Toledo Edison has sufficient experience and capability to design and con-struct the Davis-Besse Station and will have experienced and thorcughly trained personnel to operate the station at completion of construction. 1.8 IDENTIFICATION OF CONTRACIORS B&W will design and supply the nuclear steam system, core flooding systems, reactor control and protection system, feedwater control system and related reactor auxiliary systems. B&W vill also provide the majority of the in-struction for the station staff personnel. The technical qualifications of B&W are given in Appendix 1A. The turbine generator will be supplied by the General Electric Company. Bechtel Corporation and its affiliate, Bechtel Company, have been engaged as the architect-engineer for this project and as such will perfom engi- . neering and design work for the balance of plant equipment, structure and systems not included in the B&W scope of supply. Bechtel vill prepare specifications, subject to Toledo Edison's approval, for all material, equip-ment and construction contracts and will assist in evaluation of bids. Toledo Edison vill do all purchasing for the project. Bechtel will provide quality assurance assistance to Toledo Edison and has also been engaged to perform construction management services during construction. The technical qualifications of Bechtel are given in Appendix 1A. The firm of Woodward-Clyde & Associates has beea retained as a consultant to provide geotechnical services for the project. These services cover geology, siesmology, foundation design and groundwater conditions. The Great Lakes Research Division of the Institute of Science and Technology, The University of Michigan, has been retained for limnology studies of the site area. This includes lake levels, lake characteristics, lake ecology and effect of usage of lake waters. Dr. J. C. Ayers is responsible for conducting this work. The Travelers Research Corporation has been retained as a consultant in the area of meteorology. 19 CONCLUSIONS On the basis of the infomation presented in this Preliminary Safety Analysis Report, the Davis-Besse Station vill be designed, constructed and operated without undue risk to the health and safety of the public. 67 1 36

s J LA ST. DETROITo f MICHIGAN

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ML19319C280

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1.1 INTRODUCTION

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ML19319C280

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SUMMARY

1.1 INTRODUCTION

1.1 INTRODUCTION

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