ML19309C538

Chapter 1 to TMI-1 PSAR, Introduction & Summary. Includes Revisions 1-11

ML19309C538
Person / Time
Site:	Three Mile Island
Issue date:	05/01/1967
From:	JERSEY CENTRAL POWER & LIGHT CO., METROPOLITAN EDISON CO.
To:
References
NUDOCS 8004080729
Download: ML19309C538 (53)

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TABLE OF CONTENTS Section Page 1 INTRODUCTION 1ND

SUMMARY

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1.1 INTRODUCTION

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1.2 DESIGN HIGHLIGHTS 1-2 1.2.1 SITE CHARACTERISTICS , 1-2 1.2.2 POWER LEVEL l-2 1.2 3 PEAK SPECIFIC POWER LEVEL l-2 1.2.h REACTOR BUILDING SYSTEM l-2 1.2 5 ENGINEERED SAFEGUARDS 1-2 1.2.6 ELECTRICAL SYSTEMS AND EMERGENCY POWER 1-3 1.2.7 ONCE-THROUGH STEAM GENERATORS 1-k 1.3 TABULAR CHARACTERISTICS 1-k 1.h PRINCIPAL DESIGN CRITERIA

- 1-7 1.h.1 CRITERION 1 1-7 1.h.2 CRITERION 2 1-9 1.h.3 CRITERION 3 1-9 1.h.h CRITERION h 1-10 1.h.5 CRITERION 5 1-10 1.h.6 CRITERION 6 1-11 1.h.7 CRITERION 7 1-12 1.h.8 CRITERION 8

, 1-12 1.h.9 CRITERION 9 ^

1-13 1.h.10 CRITERION 10 ,1-lb 2-1  !

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1.4.11 CRITERION 11 1-1h 1.k.12 CRITERION 12 1-15 1.4.13 CRITERION 13 1-16 1.h.14 CRITERION lh 1-17 1.4.15 CRITERION 15 1-17 1.4.16 CRITERION 16 1-18

-1.h.17 CRITERION 17 1-19 1.h.18 CRITERION 18 1-20 1.h.19 CRITERION 19 l-21 1.h.20 CRITERION 20 1-21 1.h.21 CRITERION 21 1-22 1.4.22 CRITERICN 22 1-22 1.4.23 CRITERION 23 1-23 1.h.2k CRITERION 2h 1-23 1.h.25 CRITERION 25 1-2h  !

1.k.26 CRITERI0E 76 1-2h 1.h.27 CRITERION 27 1-25 15 RESEARCH AND DEVELOPMENT REQUIREMENTS 1-25 1.5 1 ONCE-IRROUGH STEAM GENERATOR TEST 1-25 152 CONTROL RCD DRIVE LINE TEST 1-26 1.5.3 SELF-POERED DETECTOR TESTS 1-26 1.5.h THERMAL AND HYDRAULIC PROGRAMS 1-26 1.6 IDENTIFICATION OF AGENTS AND CONTRACTORS 1-27 l l

17 CONCLUSIONS 1-28  !

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LIST OF TABLES Table No. Title Page 1-1 Engineered Safeguards 1-30 1-2 Canparison of Design Parameters 1-31 i

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LIST OF FIGURE Figure No. Title 1-1 Service Area 1-2 Equipment List 1-3 General Arrangement Floor Plan Elev. 281'-0" l-h General Arrangement Floor Plan Elev. 305'-0" 1-5 General Arrangement Floor Plan Elev. 329'-0" l-6 General Arrangement Floor Plan Elev. 3k6'-0" l-7 General Arrangement Floor Plan Elev. 36h'-0" l-8 General Cross Section A-A is B-B l-9 General Cross Section C-C 1-10 General Cross Section D-D l-ll Plot Plan 1-12 Extended Plot Plan 1-13 Engineering Organizational Chart t

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1 INTRODUCTION AND

1.1 INTRODUCTION

This Preliminary Safety .*nalysis Report is submitted in support of Metro-politan Edison Capany's application for a construction permit and facility license for the Three Mile Island Nuclear Station to be located on Three Mile Island near the east shore of the Susquehanna River in Dauphin County, Pennsylvania. The station location is shown on Metropolitan Edison's Service Area Map, Figure 1-1.

The generating unit vill operate initially at core power levels up to 2452 MWt which corresponds to a gross electrical output of about 8h5 MWe. All physics and core themal hydraulics infomation in this report is based upon the reference core design of 2h52 MWt. It is expected that the unit will be capable of an ultimate output of 2551 MWt (including 16 MWt contribution fra reactor coolant punps), corresponding to a gross electrical capability of about 871 MWe. Site parameters , principal structures , engineered safeguards ,

and certain hypothetical accidents are evaluated for the expected ultimate core output of 2535 MWt.

The nuclear steam supply system is a pressurized water reactor type similar to systems operating or under construction. It uses chemical shim and con-trol rods for reactivity control and generates steam with a small amount of sq trheat in once-through steam generators. The nuclear steam supply system anc first fuel core vill be supplied by The Babcock & Wilcox Capany.

Construction is scheduled for empletion in time for fuel leading to begin by December 1,1970, and for camercial operation by May 1,1971. To meet this schedule, construction of the Reactor Building is to begin not later than February 1,1968.

The general arrangement of major equipnent and structures , includin6 the Re-actor, Auxiliary, and Turbine Buildings , is shown on Figures 1-2 through 1-12.

The organization of this report follows as closely as possible the AEC's

" Guide" announced in the Federal Register on August 16, 1966. Every attempt has been made in this report to be empletely responsive to that guide, to the proposed AEC design criteria, and to all known pertinent q.uestions asked of other applicants up until the time of this writing.

This report presents descriptive material and analyses of a " reference-design."

As the station design progresses frm conceptual design to final detailed design, the station description and analyses will be subject to change and refinement.

Metropolitan Edison Capany is fully responsible for the complete safety and adequacy of the st tion. Aid in the design, construction, testing, and start-up of the unit vill be supplied principally by Gilbert Associatet, Inc. , United Engineers & Constructors Inc. , and The Babcock & Wilcox Capany. Assistance shall also be rendered by other consultants and suppliers as may .be required.

The technical qualifications of Met-Ed, Babcock & Wilcox, Gilbert Asaociates ,

and United Engineers are outlined in Appendix 1A.

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1.2 DESIGN HIGHLIGHTS 1.2.1 SITE CHARACTERISTICS I

The site is characterized by a 2,000 foot exclusion radius ; isolation frcza l nearby population centers; scund bedrock foundation for structures; an

abundant supply of cooling vater; an ample supply of emergency power; and l favorable conditions of hydrology, geology, seismology, and meteorology. ,

1.2.2 POWER LEVEL Initially licensed power for the reactor core is proposed at 2,k52 Wt, and core performance analyses in this report are based on this initial power level. Operating confi n ation of reactor core parameters is expected to support an ultimate core power level of 2,535 MWt, and the Station vill be designed to operate at this output. The analyses of accidents that could release fisMon products to the environment have been evaluated on the basis of 2,535 MWt. An additional 16 MWt will be available to the cycle from the contribution of the reactor coolant pumps, resulting in a gross electrical output of about 8h5 MW at initially licensed power and 871 MW ultimately.

1.2.3 PEAK SPECIFIC POWER LEVEL The peak specific power level in the fuel for initial operation at 2,452 MWt results in a maximum thermal output of 17 5 kw per ft of fuel rod. This value is ccuparable with other reactors of this size and therefore does not represent an extrapolation of technolcgy. This comparison may be seen in the infomation presented in Table 1-2.

1.2.h REACTOR BUILDING SYSTEM The system required to contain the Maximum Hypothetical Accident consists of the Reactor Building envelope and the engineered safeguards.

The prestressed, post-tensioned concrete Reactor Building is similar in design to the containment buildings for the Turkey Point Plant (Docket Nos. 50-250 and 251), the Palisades Plant (Docket No. 50-225), the Point Beach Plant (Docket No. 50-266), and the Oconee Nuclear Station (Docket Nos.

50-269, 50-270, and 90-2871. Several of the engineered safeguards are similar l2 to tnose plants and Three Mile Island presents neither uncczumon solutions to engineering problems nor significant extrapolations in technology.

1.2 5 ENGINEERED SAFEGUARDS Engineered safeguards are employed to reduce the potential radiation dose to the generkl public frem the Maximum Hypothetical Accident to less than the guideline values of 10 CFR 100. This is accomplished by automatic isolation of all reactor building fluid penetrations that are not required I for limiting the consequences of the accident, thus eliminating potential l leakage paths. Long-tem potential releases folleving the accident are

( reduced by rapidly decreasing the reactor building pressure to near atmos-pheric within 2h br, thereby reducing the driving potential for fission

• product escape. *

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r In addition, the engineered safeguards vill prevent core meltdown should the vorst postulated loss-of-coolant accident occur. This is accanplished by I injection core flooding systems of large-capacity, parts of which are contin-uously operated for nomal purposes and are therefore imediately available for energency duty. These systems, coupled with the themal, hydraulic, and blevdown diaracteristics of this reactor, vill reliably prevent metal-water reactions.

The engineered safeguards equipnent of the nuclear unit, along with the nomal operating modes, are as follows:

a. High pressure injection - nomally operates as part of the makeup and I purification system.  !

b. Coie flooding system

c. Low-pressure injection - operates for shutdown cooling as part of the decay heat removal system.

d. Reactor building sprsy system

e. Reactor building cooling system

f. Reactor building isolation system including penetration pressur-ization system and fluid block systci - operates on test or accident signal.

O Table 1-1 lists equi; ment supplied for tP e engineered safeguards.

1.2.6 ELECTRICAL SYSTEMS AND D4ERGENCY POWER The Three Mile Island Nuclear Station vill have the following sources 5 of electric power:

a. Three 230 kv transmissio'n lines terminating at the station atom two different directions.

b. Its own generator which vill continue to supply its aux-iliary loads upon a trip separating the substation from the transmission system.

c. Two quick-starting 2850 kv diesel-generator units connected to the safeguards busses.

Within the station there vill be multiple redundant busses and ties supplying power to loads, instruments and controls. The engineered safeguards are supplied from two separate safeguard power busses, each of which can be supplied from any of the principal sources of power. ,

i The sources of pcver and associated electrical equipment will'. insure safe functioning of the station and its engineered safeguards. 4 0000 017 1-3 (Revised 12-22-67)

7 1.2 7 ONCE-THROUGH STEAM GENERATORS The steam generators are of a new design based ou extensive research, develop-ment, and experimental work on boiling heat transfer performed by B&W over the past 10 years. Each generator is a vertical shell-and-tube, counterflow 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, and converted to steam and superheated in a single pass through the generator. The basic design parameters, such as feedvater heating, boiling length, superheat length, and perfor=ance characteristics, have been confirmed by testing of a full-length 7-tube unit and a 37-tube unit. Tests are continu-ing to provide additional data in these design areas for the 37-tube test unit. 3 In addition, testing vill continue with one 19-tube full-length test unit.

With the once-through design, natural circulation flow is adequate to remove full decay heat without the use of reactor coolant pumps. Thus, with total loss of pumps, the fuel vill not reach departure from nucleate boiling.

1.3 TABULAR CHARACTERISTICS Table 1-2 is a comparative list of important design and operating character-istics of the Met-Ed Three Mile Island Nuclear Station, Oconee Nuclear Station l Units 1 and 2 (Duke Power Company), Turkey Point Units 3 and 4 (Florida Pcver and Light Company), and Indian Point Station Unit 2 (Consolidated Edison Com-l pany of New York, Inc.). The design and operating parameters of the Oconee, l Turkey Point, and Indian Point stations are close to those of the Met-Ed facility. Oconee Units 1 and 2 each have the same rated power as the Met-Ed facility and are near-duplicates in other respects. The data in Table 1-2 represent infor=ation presented in available station descriptions, and Safety ,

Analysis Reports submitted for licensing.

The design of each of these stations is based ca information developed from operation of cc=mercial and prototype pressurized water reactors over a number

, of years. The Met-Ed design is based on this existing power reactor technology I

and has not been extended beyond the boundaries of known information or oper-ating experience.

The similarities and differences of the features of the reactor stations listed in Table 1-2 arc 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 l Most of the parameters listed in this section are similar for each reactor facility, but differ according to the thermal power level. (Note the same design basis for the Oconee and the Three Mile Island units.) The differ-ences in pcVer level are reflected chiefly in the total heat output, core size (fuel leading), coolant flow rate, and total heat transfer surface.

They amount only to a scaling down of the parameters above for a decrease in the thermal reactor power level, and do not alter the safety-related characteristics of the reactors. The Departure from Nucleate Boiling (Il 1h (Revised 11-6-67) 0000 0iB

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Ratio (DdBR) and the maxi =um ratio of peak-to-average total beat input

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per fuel rod (FAh nue.) are representative of a more conservative design for Met-Ed than for the other reactors presented. These comparisons are discussed in detail in 3.2.3.2.

Item 2. Core Mechanical Design Parameters The dimensions , =aterials , and technology for each of these reactors are similar. (Note the same design basis for the Oconee and the Three Mile Island units.) This unifomity is again due to optimization of the op-terating parameters for this type of reactor, and differences are related to the power levels.

There are also small differences in the mechanical assembly of the fuel rods and the number of control rods used in the individual reactors. The increased number of control rods in the Met-Ed reactor provides for maneu-verability and flexibility of operation. Met-Ed utilizes a canned fuel assembly which provides structural integrity and protection of the fuel rods against damage during fuel handling operations.

I tem 3 Preliminary Nuclear Design Data Since these reactors have essentially the same core geometrical configura-tion, the fuel loading differs by an amount proportional to the physical size of the reactor core. Note that the design data for the Oconee and the Three Mile Island units are the same except for the first-cycled burnup, feed enrichments , control characteristics , total red vorth, and boron con-centrations. These differencer reflect the greater first-cycle burnup of Met-Ed over Oconee. Oconee Una .1 has a first-cycle fuel sharing program with Unit 2, which requires a lower first-cycle enrichment for Unit 1.

The basis of the 2.97 (H2 0/U) is the ratio of the water molecules to the atoms 2 of U metal in the fuel assembly envelope. This value of the water to U metal -

volu=e ratio, consistent with the entry for Turkey Point Units 3 or k, is 3.53.

Each core has a three-region fuel loading, but differs in the fuel burnup ratio that is to be used.

The Met-Ed reactor design offers about 2 5 per cent greater reactivity control in the control rods. This is also reflected in the lesser concentration of boron that is required to control the reactivity over the lifetime of the reactor core. Some slight differences are noted in reactivity coefficients.

Item h. Princiet_1 Design Parameters of the Reactor Coolant System i

Most of the features in this section are directly related to material properties and the amount of heat produced in the reactor core. Note that the Three Mile' Island and the Oconee units are identical. The para-meters are scaled in proportion to the power of the reactor. The major difference between these units and Indian Point No. 2 and Turkey Point Nos. 2 and 3 is the number of coolant loops required to remove the heat products.

l For Three Mile Island and Oconee, only two loops are required since once-  !

through steam generators are used instead of the U-tubes-in-shell design. *I

f. ,,The greater cooling capacity of these steam generators pemits a reduc- l tion /La the number of cooling loops for an equivalent amount of heat 1 removw -  :

0000 019 1-5 (Revised 10-2-67)  ;

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[ 7 Item 5 Reactor Coolant System - Code Recuirements The Three Mile Island and the Oconee units are identical. Ccde requirements O

for the shell side of the steam genereter confer = to the ASME III Class A ,

specification. This is considered to be a contribution to the safety of the vessel since it enhances the integrity because of the most stringent ASME III Class A design, material, and quality control requirements.

Item 6. Princital Design Para =eters of the Reactor Vessel The Three Mile Island and the Oconee units are identical. These vessel designs are characterized by a thinner thermal shield and a relatively larger diameter.

The larger diameter provides for additional vater between the edge of the core and the vessel, which leads to additional neutron attenuation.

Item 7 Princical Design Features of the Steam Generators The steam generators in the Three Mile Island and the Oconee fccilities are the sa=e. They are basically different frc= Indian Point Unit 2 and Turkey Point since they are once-through design and incorporate an integral superheat section.

Item 8. Princical Design Parameters of the Reacter Coolant Pwsts The Three Mile Island and the Oconee designs are identical. In each specific para =eter the relative number or size is in proportion to the total a= cunt of heat removed frem the core. The Three Mile Island pumps have higher head and hersepower requirements than the Turkey Point and Indian Point Unit 2 have for approximately the same flow because of the increased flow lesses of the once-through steam generators and the use of only two reactor coolant loops.

Item 9 Priueical Design Parameters of the Reactor Coolant Pining The Three Mile Island and the Oconee piping designs are the sa=e. They both utilize carben steel clad with stainless steel.

Item 10. Reactor Building Parameters

The Three Mile Island reacter building is basically the same design and con-i struction as the Oconee and Turkey Point units. The differences are physical dL=ensions, a= cunt of concrete shielding needed, and d.esign incident pressures, l which are a direct result of station layout, engineered safeguards, system l capacities, and site location. The reacter building design and shielding offer satisfactory protection to the surrounding population in case of an acci-dent and during normal operation of the generating units.

e l Item 11. Engineered Safeguards l

l l Engineered saieguards are generally similar, but Oconee includes a penetration l rocs ventilation system for each unit. The Ocenee and Turkey foint units do l not use sedium thicsulfate injection in the reactor building oiray for iodine l removal because of the differences in design and site characteristics. l3 O

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1.4 PRINCIPAL DESIGN CRITERIA The Three Mile Island Nuclear Station vill be designed with the intent to meet th9 ST General Design Criteria for Nuclear Power Plant Construction 2 Permitslll proposed by the AEC in Nov9myer 1965 and those of the 70 General Design Criteria proposed in July 1967 W , which are appideable to this station.

The principal safety features that meet each criterion are summarized herein for the original 27 General Design Criteria and in " Supplement #1" (question 1.1) for the proposed 70 criteria. In the discussion of nach criterion, reference is made to sections of this report where more detailed information is presented. ,

l.h.1 CRITERION 1 Those features of reactor facilities which are essential to the prevention of accidents or to the mitigation of their consequences must be designed, fabricated, and erected tot

a. Quality standards that reflect the importance of the safety func-tion to be perfomed. It should be recognized, in this respect, that design codes 5.anmonly used for nonnuclear applications may not be adequate.

b. Perfomance standards that vill enable the' facility to withstand, without loss of the capability to protect the public, the addi-tional forces imposed by the most severe earthquakes , flooding conditions, vinds, ice, and other natural phenomena anticipated at the proposed site.

Ansver:

The integrity of systems, structures , and emponents essential to accident prevention and to mitigation of accident censequences has been considered in the design evaluations. These systems , structures , and emponents are:

1. Fuel assemblies.

2. ReTetor vessel internals.

3. Reactor coolant system,

h. Reactor instrianentation, controls , and protection systems.

5 Engineered safeguards.

6. Radioactive materials handling systems.

7. Reactor building

8. Electric power sources.

a. Quality Standards The fuel assemblies are designed to maintain their integrity when srbjected to the mechanical, hydraulic, and themal stresses resulting frca anticipated operating conditions during their design life. The design is based on technology which has proved successful in existing nuclear power plants and is substantiated by'te%t data. The fuel assemblies vill be manufactured to high quality standards and subjected to a series of rigorots tests during fabrication. (Section 3.2.h.2) l (1) The criteria as proposed by the AEC in its press release H-252 of November 22, 1965 .

( , (2) (l"h,e criteria as proposed by the AEC in its press release K-172 of

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Ca ponents and piping in the reactor coolant system are designed for a pressure of 2,500 psig at a temperature of 650 F. The nminal operating conditions of 2,185 psig and 579 F allev an adequate margin for normal load changes and operating transients.

The reactor coolant system is designed to meet applicable portions of the following principal codes: (Section k.1.5)

Piping and Valves - ASA B31.1 (Piping Code) including nuclear cases. .

Punp Casing - ASME Boiler and Pressure Vessel Code,Section III.

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

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

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

Quality control, inspection, and system testing vill insure integrity of the reactor coolant system. (Sections h.1 and k.h)

The fast neutron exposure (3 x 10 19 nyt) of the reactor vessel results in a maximum NDTT of not more than 260 F at the end of Station service life. Operating procedures for the Station vill be capatible with these temperature limitations. (Section h.1.4)

The instrumentation, reactor control system, and protection systems vill be fabricated using high quality empenents and workmanship. All cmponents vill be selected on the basis of reliability, durability under service conditions , and proven

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application. In the absence of applicable industry standards and codes regulating quality in this type of equipment, quality standards ce established based on extensive experience, study, and testing. (Section 7)

Engineered safeguards , i.e. , emerger;g/ injection, core flooding system, the Reactor Building spray system, the Reactor Building isolation system, and the Reactor Building emergency cc;11ng units, are designed and vill be fabricated to high quality standards and tested to insure proper operability. All piping in these systems is designed to the applicable ASA Code for l pressure piping. (Section 6)

Radioactive material handling systems are designed and vill be fabricated in accordance with the applicable design codes listed in the introduction to Section 9 Pressure vessels are designed to either ASME Boiler and Pressure Vessel Code Sections III or VIII, as applicable.

The Reactor Building is designed and vill be constructed in accordance with applicable sections of appropriate ACI a.ad ASTM codes and specifications as well as criteria described in Sec- ,

tion 5 .

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Three normal sources of electric power vill supply the electri-cal load requirements as follows : (a) station auxiliary trans-fomer No.1, (b) station auxiliary transfomer No. 2, and (c) engineered safeguards transfomer. Four redundant battery-backed busses vill be provided for vital instrumentation and control. Electrical equipment vill be purchased and tested to stringent requirements for reliability and quality, including appropriate NEMA, ASA and IEEE electrical standards. (Section 8)

b. Perfomance Standards All equipment and structures having a safety function vill be designed, constructed, operated, and maintained without loss of capability to protect the public under all environmental conditions anticipated at the site.

l.h.2 CRITERION 2 Provisions must be included to limit the extent and the consequences of credible chemical reactions that could cause or materially augment the release of significant amounts of fission products from the facility.

s Answer:

Eter6ency injection coolant is provided to maintain the core sufficiently ,

covered to prevent core melting for the ccanplete range of postulated '

reactor coolant system rupture sizes up to the maximian size of a 36 in.

ID pipe. In the process of cooling the core, the metal-water reaction is limited to an insignificant amount. (Section 1h)

High pressure injection with a capacity of 1,000 gpn vi:1 pump water to the reactor coolant system at all pressures up to full design pressure. l This system is primarily effective early in the accident while the reactor cooJ ant system pressure is above 100 psig. (Section 6) l A core flooding system is supplied to cool the core at intermediate to low pressures. This system consists of two independent core flooding tanks, each of which is connected to a different reactor vessel injection  ;

nozzle.

j Low pressure injection provides 9,000 gpn to cool the core when it is 1 partially or totally uncovered and the reactor coolant pressure has dropped below 100 psig. (Section 6) 1.h.3 CRITERION 3 l

Protection must be provided against possibilities for damage of the safe-guarding features of the facility by missiles generated thrbugh equipment failures inside the containment. '

Answer :

Frotective valls and slabs , local missile shielding, or restraining devices will be provided to protect the Reactor Building liner plate and engineered j safeguards within the Reactor Building against dama6e from missiles. gene-

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rated by equipment failures. The cencrete enclosing the reactor coolant j system serves as radiation shielding and an effective barrier against missiles. Local missile barriers will be provided for control red drive mechanis=s . (Sections 4 and 51.2 7.1) l For those parts of the safeguards susceptible to missile damage, redundant equipment is provided to insure required operation. (Section 6) 1.4.4 CRITERION 4 The reactor must be designed to accamodate, wit'hout fuel failure or pri-mary system damage, deviations frca steady state norm that might be occasioned by abnormel yet anticipated transient events such as tripping of the turbine-generator and loss of pover to the reactor recirculation system pumps. ,

Ansver:

The reactor is designed with a margin above normal operating conditions to acemmodate anticipated abnormal deviations frcm the steady state l operation. This margin allows for deviations of temperature, pressure, flow, reactor power, and reactor- turbine power mismatch. The reactor is operated at A constant avertie coolant temperature above 15 per cent power and has a negative power coefficient to dampen the effects of power trans-ients. The reactor control system vill maintain the reactor operating parameters within preset limits, and the reactor protection system vill shut down the reactor if normal operating limits are exceeded by preset amounts. (Sections 7.1 and 14)

The nuclear unit is shut down autcmatically by the reactor protection system if a ecmplete loss of electrical power occurs. Upon loss of external system electrical load, a reactor power reduction occurs, and the reactor continues to generate Station power needs at reduced load. The resultant reactor coolant system temperature and volume increases for both of the above are held within design limits by relieving steam through the bypass to the condenser and/or secondary system relief valves to the atmos-phere, thereby preventing excessive reactor coolant system pressures.

Arcordingly, these transients vill not produce fuel or reactor coolant system damage. (Sections 7.1 and 14.1.2.8)

The reactor coolant pumps are provided with sufficient inertia to maintain adequate flew to prevent fuel damage if power to all pumps is lost. The l criterion for core protection folloving loss-of-coolant flow is to maintain l a Departure frca Nucleate Boiling Ratio (DNBR) equal to or greater than l that at the design overpower level for initial power conditions up to and

including the maximum operating power level of 107.5 per cent power. Natural l circulation coolant flow will provide adequate core cooling after the pump energy has been dissipe.ced. (Section 14.1.2.6 and Figure 9-T) l l 1.4.5 CRITERION 5 The reactor must be designed so that power or process variable oscilla-tions or transients that could cause fuel failure or primary system damage are not possible or can be reedily suppressed.

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l Answer:

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The ability of the reactor control and protection system to control the oscillations resulting fra variation of coolant temperature within the control syster dead band and frem spatial xenon oscillations has been )

analyzed. Variations in average coolant temperature provids negatire feedback and enhance reactor stability during that portion of core life in which the moderator temperature coefficient is negative. When the coafficient is positive, rod motion vill ecmpensate for the positive feed-back. The maximum power change rate resulting frcm temperature oscilla-tions within the control system deadband has been calculated to be less than 1 per cent per minute. Since the Station has been designed to follow i l

ramp load changes of 10 per cent per minute, this is well within the capability of the control system. {

l Control flexibility with respect to xenon transients is provided by the c abination of control rods, incore instrumentation, and out-of-core instrumentation. Within control rod limits, transient xenon related to load changs is controlled by the autmatic control system. Arial, radial, or azimuthal neutron flux changes vill be detected by both out-of-core and .

incere instrumentation. Individual or groups of control rods can be posi-  !

tiened to suppress and/or correct flux changes. (Sectic? 3.2.2.2.3) '

1.h.6 CRITERION 6 Clad fuel n.ust be designed to accommodate throughout its design lifetime O all normal and abnormal modes of anticipated reactor operation, including the design overpower condition, without experiencing significant cladding 1

l failures. Unclad or vented fuels must be designed with the similar objec-  ;

tive or providing control over fission products. For unclad and vented '

solid fuels , normal and abnomal modes of anticipated reactor operation must be achieved without exceeding design release rates of fission products from the fuel over core lifetime.

Answer:

Fuel clad integrity is insured under all normal and abnomal modes of anticipated operation by avoiding clad overstressing and overheating. I The evaluation of clad stresses includes the effects of internal and )

external pressures, temperature gradients and changes , clad-fuel inter-i actions, vibrations, and earthquake effects. The free-standing clad design i prevents collapse at the end volume region of the fuel rod and provides i sufficient radial and end void volume to accamodate clad-fuel interactions and internal gas pressures. (Sections 3.2.4.2) l Clad overheating is prevented by satisfying the following core themal and hydraulic criteria: (Section 3.2.3.1.1) ,

l j a. At the design overpower no fuel melting vill occur.

b. A 99 per cent confidence exists that at least 99.5 per cent of the fuel rods in the core vill be in no jeopardy of experiencing

• O a DBN during continuous operation at the design overpovet of 114 per cent.

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1.h.' CRITERION 7 The maximum reactivity worth of control rods or elements and the rates with which reactivity can be inserted must be held to values such that no single credible mechanical or electrical control system malfunction l

could cause a reactivity transient capable of damaging the primary system or causing significant fuel failure.

Answer:

l Reactivity control vill be acceplished by movement of control rods and by changes in scluble poison (boron) concentration in the reactor coolant.

Each control rod assembly consists of a cluster of 16 poison. rods. The rod

( drives and their controls vill have an inherent feature to limit overspeed in the event of malfunctions.

Approach to criticality and low power operation vill be by manual rod withdrawal. The remaining rods 'or rod croups) vill be interlocked to l pemit withdrawal on autmatic control only after the rod groups used for approach to criticality and low power operation have been .W vith-drawn. Rods used fer autmatic control vill be arranged in four groups and interlocked to prevent simultan. " withdrawal of more than two l groups. That is, simultaneous withdrawal of two automatic groups vill be pemitted over approximately .the first 25 per cent of the second rod j group stroke and the last 25 per cent of the first red group stroke.

simultaneous with-The drawalmaximum reactivity of a regulating insertion twelve rate associated rod group is 5.8 x 10-vitg Ak/k/sec.Assuming a single electrical failure occurs that invalidates the interlock and pemits the 25 control rods on automatic control to be vithdrawn simultaneously, a maximum reactivity insertion rate of 2.3 x 10-g Ak/k/see could result.

Reactivity transients of this magnitude have been analy::ed, and the resultant power transients vill not produce reactor coolant system or fuel failure.

(Section 14.1.2.3) l A reduction in the reactor coolant soluble poison concentration vill require l

operator initiation, and vill be prohibited by interlecks until the control rods are in an acceptable pattern for dilution. A second safety feature vill physi'cally limit the maximum rate at which dilution water can be added to the system. A third safety measure vill consist of a re.'.ay which vill limit the total time of dilution. The maxi =um reactivity insertion rates fra moderator dilution vill be 7 0 x 10-DA k/k/sec. These rates are not sufficient to produce damage to either the fuel or reactor coolant system.

i l.h.8 CRITERION 8 Reactivity shutdown capability must be provided to make and hold the core i suberitical frm any credible operating condition with any one control l element at its position of highest reactivity.

G o

300326 1-12

() Answer:

The reactor is designed with the capability of providing a shutdcwn margin of at least 1 per cent ak/k with the single mest reactive control rod fully with-drawn at any point in core life with the reactor at a hot zero pcwer condition.

The minimum hot shutdown margin of 2.1 per cent ak/k occurs at the end of life. l3 Reactor suberitical =argin is maintained during cooldevn by changes in soluble poison concentration. The rate of reactivity compensat4cn from boren addition is greater than the reactivity change associated with the maximum allowable reactor cooldcun rate of 100 F per hour. Thus , suberiticality is assur.ed dur-ing cooldown with the mest reactive control red totally unavailable. (Section 3.2.2.1) 1.k.9 CRITERION 9 .

Backup reactivity shutdcwn capability =ust be provided that is independent of normal reactivity control provisions. This system must have the capability to shut down the reactor frc= any operating condition.

Answer:

Soluble poisen addition will provide an independent backup to the control rods for reactivity shutdown. Poison addition will be acecmplished using the =ake-up and purification 'syste=. There are two =akeup pumps to insure flow avail-ability under all credible operating conditions. These pu=ps take suction

(~~)g from the borated water storage tank, which contains water with 2,270 ppm boren,

(_ or from the makeup tank. In the latter case, a solution containing 8,750 ppm bcron is supplied to the =akeup tank from a sixing tank. Two transfer pumps are provided. (Section 9 1)

The makeup pumps and the two sources of concentrated boron solution insure the capability of being able to shut down the reactor without any control rods frem any operating condition. The following table demonstrates the capability of shutdown without centrol rods for two modes of makeup and purification syste= l3 operation. '

Soluble Poison S,hutdown Capability Negative Reactivity Time to Shut Down From 100%

Feed Insertion Rate, Full Pcwer to Hot Zero Power Concen- Feed 5 ok/k/rinute Condition!1), minutes tration, Flow Rate, ppm baron gpm BOL EOL BOL EOL -

8,750 20 0.0179 0.0217 67 106 3 L,121 70(2) 0.022h 0.0353 Sh 65 3,196 lh0(3) 0.0278 0.05L3 h3 h2 O V ":: , .

0000 027

~

'-13 (Revised 11-6-67) 1

(1) Reactivity balance on Doppler and moderator equal to 1.2% ak/k for 3 BOL and 2.3% ak/k for EOL.

(2) Makeup to makeup tank at 20 gpm of 3,750 ppm boren from boric acid h

six tank plus 50 gym at 2,270 ppm boren from storage tanks.

(3) Makeup to makeup tank at 20 gpm of 8,750 ppm bcron fro = boric acid l3 mix tank plus 120 gpm at 2,270 ppm boren frem storage tanks.

1.h.10. CRITERICN 10 Heat removal systems must be provided which are capable of acccm=odating core decay heat under all anticipated abnor=al and credible accident conditions, such as isolation frem the main condenser and ecmplete or partial loss of pri-

=ary coolant from the reactor.

Ansver:

Reactor decay heat vill be removed through the steam generators until the re-actor coolant system is cooled to 250 F. Ste u generated by decay heat vill supply the steam-driven feedvater pu=p turbine and can also be ven;ed to atmo-sphere and/or bypassed to the condenser. The steam generators are supplied feedvater frem either the main steam-driven feedvater pumps, which can be op-erated at a reduced flev rate for decay heat removal, or frem a steam-driven emergency feed pump sized at 5 per cent of full feedvater ficv.

The main feedvater pumps supply water contained in the feedvater train and the condensate storage tank to the steam generators. The emergency feed pu=p takes suction frem the condenser hotvell and the condensate storage tank. These sources provide at least 200,000 gal of water storage which is sufficient for /

decay heat removal for about one day after reactor shutdown with the condenser '

isolated. The condenser is normally available so that water inventory is not depleted. (Section 10)

Without use of reactor coolant pumps , decay heat will be removed by natural circulation through the reactor coolant system. (Section 14.1.2.8)

Under conditions of complete or partial loss-of-coolant frc= the reactor, decay heat will be removed from the core by coolant supplied by the emergency injection coolant systems. The source of injection water vill be the borated water storage tank. When this source is exhausted, the low pressure injection pumps vill take suction frem the Reacter Building su=p. The return flow is cooled and pumped te the reactor vessel to continue core cooling. This sys-tem contains redundancy of equipment to insure availability of flow when re-quired. If complete less of external electric power occurs, on-site sources supply sufficient electric power for all engineered safeguards and cooling water systems. (Section 14.2.2.3) 1.4.11 CRITERION 11 g Components of the primary coolant and containment systems must be designed

, and operated so that no substantial pressure or thermal stress will be is-l posed on the structural materials unless the temperatures are well above the nil-ductility temperatures. For ferritic =aterials of the ecolant envelope and the containment, =inimum temperatures are NDT + 60 F and NDT + 30 F, respectively. .

, s l-14 (Revised 11-6-67) 0000 028

Answer:

The reactor vessel plate material opposite the core is purchased to a specified NDTT of 10 F or less and is tested to verify confomity to specified requirenents. (Section 4.2.5)

The end-of-Station-life-NDTT value of the reactor vessel opposite the core vin be not more than 260 F. Station operating procedures will be estab-lished to limit the operating pressure to 20 per cent of the design pressure when the reactor coolant system temperature is below NITIT plus 60 F through-out Station life. Surveillance specimens of the reactor vessel shen section material win be instaued between the core and inside vall of the vessel shell to monitor the NDTT of the vessel material during operating lifetime.

(Section k.1.k)

The reactor vessel material is protected frca excessive radiation damage by '

coolant water annuli between the core and the reactor vessel. The thickness of these annuli limits the total fast flux grgter than 1 Mev incident on the reactor vessel van to an nyt value of 3 x 10 in ho years at an 80 per cent station capacity factor. The themal shield contributes to a further reduc-tion in vessel material radiation damage. (SwU ' l.k)

The Reactor Buildisg liner is enclosed within the Reactor Building and thus will not be exposed to the temperature extremes of the environs.

The Reactor Building ambient temperature during plant operation vill be between 100 and no F which is expected to be ven above the NDT temperature

+30 F for the liner plate. The liner plate is completely enclosed by the thick concrete vans, slab, and roof of the Reactor Building, and vill thus not be subject to sudden variations due to changes in external temperatures. In addition, the bottom liner plate is protected by a minimum thickness of 21 in. of concrete. The penetrations vill be made of material which exhibits by test a transition temperature at least 30 F below the minimum service metal temperature. For the purpose of determining minimum service metal temperature the lowest one' day mean temperature vill be assumed as -5F.

1.k.12 CRITERION 12 Capability for control rod insertion under abnomal conditions must be provided.

Answer:

Control rods will provide the norsal means for changing reactivity to shut down to a hot suberitical condition. They may be inserted ' independently ,

of the normal reactor control system by the reactor protection system or by manual means. Both modes of insertion override reactor control system signals by interrupting power to the rod drives. Without power the control rods insert into the core by gravity. Soluble poison is added to maintain suberiticality fr:xa a hot to a cold zero power condition.

e The principal safety criteria for the control rod drives are:

O a. No single failure in +1e drive shan result in the loss of safety function. p l-15 0000 029

.n. g',n..n.

b. Trip action shall not require power, and no single failure or chain of failures shan prevent trip action to more than one mechanism.

c. The trip cccmand shall override all other ccx::mands. Trip action shan be nonreversible.

The reactor vessel, reactor vessel supports , reactor vessel internals ,

fuel assemblies, control rods , and the control rod drive are all designed to resist, without loss of function, the effects of seismic loadings established by the seismological analysis of the site.

The control rod is never withdrawn ecmpletely frem the fuel assembly. The guide structure is oriented with respect to the fuel assembly by a ccumnon grid structure which maintains full stroke control rod guidance into the fuel assembly. The drive line is designed and vill be tested to be fully operable under conditions of the maximum misalignment specified. (Section 3.3 3.4.1) 1.k.13 CRITERION 13 The reactor facility =ust be provided with a control room frcan which all actions can be controlled or monitored as necessary to maintain safe operational status of the plant at all times. The control recun must be provided with adequate protection to pennit occupancy under the condi-tions described in Criterion 17, and with the means to shut down the plant and maintain it in a safe condition if such accident were to be experienced.

Answer:

The reactor unit vill be controlled frcm a separate control panel located in the control rocm. The control rom is designed to permit continuous occupancy following an accident. (Section 7.k.5)

All controls and instrumentation required to monitor and operate the re-actor and electric power generating equipment vill be located within the control room. This includes indication cf power level, process variables such as temperatures, pressures , and flows , valve positions , and control red position.

All engineered safeguards equipment vill be controlled and monitored frca the control rocan. The status of all dynamic equipnent (pumps , valves , etc.

as ven as pertinent pressures, temperatures , and flows vill be displayed.

The station radiation monitoring sys'.em vill have instrumentation readouts displayed in the control rocm.

l During MHA conditions the concrete Reactor Building and control roca valls I

and roof provide adequate protection against direct radiation to control l room personnel. Control roca personnel on eight hour shifts during a 90 i

day period following the MHA vould not receive an integrated direct dose l

in excess of 3 rem, frcan all sources of radiation which is approximately equal to 'the calendar quarter dose pennitted in 10 CFR 20.

O o 1-16 0000 030 1

The control ro m is provided with independent ventilation and filtration systems to control ingress of airborne contaminants escaping fra the Reactor Building. (Section 9.8.2) The activity in the control rom during any 90 day occupancy will not exceed the limits specified in 10 CFR 20.

l.h.14 CRITERION lh Means must be 1neluded in the control room to show the relative reactivity status of the reactor such as position indication of mechanical rods or concentrations of chemical poisons.

Answer:

The position of each control rod vill be displayed in the control rom.

The reactivity status of soluble poison will be indicated by the position of the control rods. The soluble poison ccacentration vill be adjusted to be consistent with specified rod patterns and control rod group posi-tion. Accordingly, continuous indication of soluble poison concentration will not be required. The operator vill receive results of laboratory analyses of the soluble poison concentration. (Section 7) 1.h.15 CRITERION 15 A reliable reactor protection system must be provided to autmatically initiate appropriate action to prevent safety limits from being exceeded.

Capability must be provided for testing functional operability of the system and for detemining that no component or circuit failure has occurred. For irstruments and control systems in vital areas where the potential consequences of failure require redundancy, the redundant channels must be independent and must be capable of being tested to de-temine that they remain independent. Sufficient redutdancy must be pro-vided that failure er removal from service of a single caponent or channel vill not inhibit necessary safety action when required. These criteria should, where applicable, be satisfied by the instrumentation associated with containment closure and isolation systems, afterheat renoval and core cooling systems, systems to prevent cold-slug accidents , and other vital systems, as well as the reactor nuclear and prccess safety system.

Answer:

The reactor protection system is designed to provide the features specified in this criterion. A minimum of four sensors are provided for each trip variable except startup rate. Two sensors are provided for startup rate monitoring. Reactor trip is provided when the following parameters exceed preset values:

a. Reactor power.

b. Reactor outlet temperature.

c. Reactor pressure.

d. Reactor startup rate.

0000 031

. 9 6

If a portion or an of an instrumentation cl.unel is removed frcm service, the channel assumes a tripped condition. One channel in a tripped condi-tion places the protection system in a half-tripped mode such that a trip of any one of the remaining channels causes a reactor trip.

Reactor Building isolation and engineered safeguards are initiated frcri a 3-channel system described in Section 7 The power supply for each individual channel vill be from one of the 4 re-dundant battery-backed vital busses. (Section 8) The channels are nomally energized and loss of power to two busses causes a reactor trip. (Section 7)

Provisions will be included for testing the protection systems and/or canponents under administrative control on a periodic basis. Normal testing vill include the insertion of a simulated signal to dynamically check response and perfomance of each channel's ccmponents except de-tectors. Tests of each protection system channel vill insure a high confidence level of system operability. (Section 7.1.3 5) 1.k.16 CRITERION 16 The vital instrumentation systems of Criterion 15 must be designed so that no credible canbination of circumstances can interfere with the perfomance of a safety function when it is needed. In particular, the effect of influences common to redundant channels which are intended to be independent

=ust not negate the operability of a safety system. The effects of gross disconnection of the system, loss of energy (electric power, instrument air), and adverse environment (heat from loss of instrument cooling, extreme ,

cold, fire, steam, water, etc.) must cause the system to go into its safest state (fail-safe) or be demonstrably tolerable on some other basis.

Answer:

Protection systems instrumentation is designed to operate in Reactor Build-ing ambient conditions ranging from h0 F to 1k0 F vithout adverse effects in accuracy. Reactor Building temperature vill be normally controlled in the range of 60 F to 110 F. The protection system instrumentation, exclus-ive of the neutron detectors in the Reactor Building, vill withstand the external pressure and temperature for the duration of a loss-of-coolant accident and still be operable (but subject to several per cent inaccuracy).

The out-of-core neutron detectors are designed for continuous operation in a tenperature of 175 F and a pressure of 150 psig.

Redundant instrument channels are provided for all reactor protection and engineered safeguard systems. Loss of power to each individual reactor pro-tection channel vill trip that individual channel. Loss of all instrument power vill trip the reactor protection system, thereby releasing the control rods. Engineered safeguards are nornally de-energized controls. They will be activated through redundant controls and power systems. (Sectioh 7 1) 1 Manual reactor trip is designed so that failure of the automatic reactor trip circuitry will not prchibit or negate the manual: trip. The same is true with respect to manual operation of the engineered safeguards equip-

, ment. (Section 7.1) 1-18 l (.-'*v:l.

i t

l l

i 1.h.17 CRITERION 17 The containment structure, including access openings and penetrations, must be designed and fabricated to acceannodate or dissipate without failure the i pressures and temperatures associated with the largest credible energy  ;

release including the effects of credible metal-vater or other chemical '

reactions uninhibited by active quenching systems. If part of the primary coolant system is outside the primary reactor containment, appropriate ]

safeguards must be provided for that part if necessary, to protect the health and safety of the public, in case of an accidental rupture in that part of the system. The appropriateness of safeguards such as isolation valves, additional containment, etc. , vill depend on environmental and population conditions surrounding the site.

Answer: ,

The Reactor Building, including access openings and penetrations , has a design pressure of 55 psig at 281 F. The greatest transient peak pressure, associated with a hypothetical rupture of the piping in the reactor coolant system and the effects of a credible metal-water reaction, vill not exceed .

these values.

The Reactor Building and engineered safeguarts systems have been evaluated for various cabinations of credible energy releases. The analysis accounts for systen energy, decay heat, metal-vater reactions , and the burning of the resultant hydrogen. The cooling capacity of either Reactor Building O cooling systen (see Criterion 18) is adequate to prevent over-pressurization of thc structure, and to return the Reactor Building to near atmospheric prassure within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The details of this evaluation are discussed in Section 14.2.2.3.5 The use of injection systems for core flooding vill limit the Reactor Build-ing pressure to less than the design pressure. If a metal-vater reaction is uninhibited by the active quenching systems the resultant peak Reactor Building pressure is less than the design pressure.

No lines which contain high temperature, high pressure reactor coolant penetrate the Reactor Building except the sampling lines. These small sampling lines are normally isolated by two valves in series. Therefore, it is only during a sampling operation that a line failure would require operator action to prevent escape of coolant external to the Reactor Building. This is a procedure that the operator would normally perform.

The makeup and purification system diverts a small amount of reactor coolant outside of the Reactor Building. This high pressure and high temperature coolant is cooled before it leaves the Reactor Building. Lines serving this function contain isolation valles that can be closed to prevent uncontrolled release of reactor coolant in the event a line fails external to the Reactor Building. The letdown coolers are supplied with water from the intermediate cooling system. Any leakage of reactor coolant through the letdown coolers vill be into this system rather than to the environment. The intermediate cooling system is monitored to detect leakage of reactor coolant. * -

O .

. ~

..o m+.. m~ 0000 133 1-19 -

i Leakage of contaminated coolant fra engineered safeguards equipment located external to the Reactor Building has been evaluated, and the '

resultant environmental consequences are well below 10 CFR 100 limits i at the site boundary, and have been included in the total accidental dose calculations.

The high pressure injection and decay heat removal systems have re-dundancy of equipent to insure availability of capacity. (Section 6.1) 1 Sme engineered safeguards systens have both a nomal and an emergency function, thereby providing nearly continuous testing of operability.

For example, one high pressure injection pump is in centinuous use for seal injection and makeup; the decay heat removal ptanps are in use for decay heat removal during each shutdown; and one Reactor Build-ing intermediate cooling system pump is in continuous use.

During normal operation the standby and operating units vill be rotated into service on a scheduled basis. In cases where separate equipment is used solely for energency conditions , such as the Reactor Building spray pumps, recirculating lines are provided, and instrumentation is installed to provide means for conducting tests. The equipent is located to facilitate inspection during cperation. (Section 6)

Electric motors', valves , and damper operators , which must function within the Reactor Building during accident conditions, vill operate in a steam-air atmosphere at 281 F and 55 psig.

1.4.18 CRITERION 18 Provisions must be made for the renoval of heat fra within the con-tainment structure as necessary to maintain the integrity of the structure under conditions described in Criterion 17 above. If engin-eered safeguards are needed to prevent containment vessel failure due to heat re). eased under such conditions , at least two independent systems must be provided, preferably of different principles. Backup equip ent (e.g. , water and power systems) to such engineered safeguards must also be redundant.

Answer:

Reactor Building cooling following the loss-of-coolant accifent is pro-vided by two independent systems: (1) the Resctor Building spray, and (2) the Reactor Building emergency coolers. The capability of either of these cooling systems, or both at partial capacity, is sufficient to prevent excessive Reactor Building pressure during loss-of-coolant accident conditions. ,

The Reactor Building spray system supplies 3,000 gpn frm thessimmiend water storage tank into the Reactor Building. After the borated water storage tank is emptied, recirculation fra the Reactor Building stanp begins. The nuclear services cooling water system is always in opera-tion, and therefore has continuously indicated availability.

. G s

1-20 '

0000 034

Two sets of nozzles, located in the upper portion of the Reactor Building structure, are arranged to provide a uniform spray pattern. Redundancy in both pumping and heat exchanger capacity exists. (Section 6.2)

To prevent excessive temperature rise following an accident, the Reactor Building cooling system has three emergency cooling units which reject lk heat to the nuclear services cooling water system. Pumps and heat exchangers are redundant to insure availability. (Section 6.2)

Upon loss of external sources of electric power, two of the three. diesel generators will peznit operation of all required engineered safeguards equipment. (Section 8.2.3) 1.4.19 CRITERION 19 The maximum integrated leakage from the containment structure under the conditions described in Criterica 17 above must meet the site exposure criteria set forth in 10 CFR 100. The containment structure must be designed so that the containment can be leak tested at least to design pressure conditions after empletion and installation of all penetrations, and the leakage rate measured over a suitable period to verify its con-femance with required performance. The plant must be designed for later tests at suitable pressures.

Answer:

The Reactor Building leakage rate vill be detezinined at design pressure after empletion and installation of all penetrations. The leak rate test vill verify that the maximum integrated leakage does not exceed 0.2 per cent by weight of the contained air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at the design pressure. (Section 5 1.2.2)

The environmental hazards from the maximum hypothetical accident, assum-ing the above specified maximum integrated leakage frem the Reactor Building, is within the guide line values of 10 CFR 100. (Section 14.2.2.k)

In order to maintain the specified leakage rate of the Reactor Building during the operating life of the station, a program of leak rate tests at suitable pressures will be established. Redundant stored energy systems vill be provided for pressurizing the access lock, equipnent hatch, elec-trical and piping penetration sleeves during accident conditions and for i

adding or maintaining a fluid block in lines upstream of isolation valves not connected to a completely closed systen within the contsizanent building and not required for operation in the accident mode. Electrical penetra-tions vill have air pressure blocks during accident conditions. These systems will be testable and are designed to minimize requirements for in-tegrated contairanent leak rate tests.

1.k.20 CRITERION 20 All containment structure penetrations subject to failure such as resilient i

seals and expansion bellows must be designed and constructed so that leak-O '

. fightyss can be demonstrated at design pressure at any tine throughout

' operatic ('Lifeofthereactor. .

0000 035 l-21 (Revised 12-8-67)

l i

Answer:

All Reactor Building penetrations with resilient seals or expansion bellows vill be constructed so that they may be pressurized during accident conditions and so that leak tests can be conducted at any time. (Sections 5 1.2.6.1, 5.h) 1.h.21 CRITERION 21 Sufficient normal and emergency sources of electrical power must be provided l to assure a capability for prompt shutdown and continued maintenance of j the reactor facility in a safe condition under all credible circumstances.

l Answer:

i l The design of the electrical system for this nuclear station vill provide l

four sources of electric power. These are: (1) power to the nuclear unit frcm its own generator; (2) 230 kv transmissica system; (3) 115 kv transmission system; and (h) three emergency diesel-engine generators.

l All of these power sources are large in capacity and, with associated equipment, will insure safe reliable functioning of the station.

1.h.22 CRITERION 22 Valves and their associated apparatus that are essential to the containment function =ust be redundant and so arranged that no credible combination of circumstances can interfere with their necessary functioning. Such redundant valves and associated apparatus must be independent of each other. Capability must be provided for testing functional operability of these valves and associated equipment to determine that no failure has occurred and that leakage is within acceptable limits. Redundant valves and auxiliaries must be independent. Containment closure valves must be actuated by instrumentation, control circuits , and energy sources which satisfy Criteria 15 and 16 above.

- Answer:

The isolation system closes all fluid lines (except those associated with engineered safeguards systems) penetrating the Reactor Building in the event of a less-of-ecolant accident. Reactor Building isolation occurs on r. signal of approximately k psig or by manual actuation from the control roem.

The criterion for isolation valve requirements is:

Leakage through all fluid penetrations including those serving accident-l consequence-limiting systems is to be minimized by a double barrier so i

that no single credible failure or malfunction of an active ccaponent can result in a loss of isolation or intolerable leakage. The double

barriers take the form of closed piping systems both inside and outside l the Reactor Building and various arrangements of isolation valves. (Section 1

5 2)

Fluid penetrations serving engineered safeguards systems also meet this j criterion, but the actuators can be manually operated frem the control I roca for test purposes.

t

. 1-22 gh s . - -

The control circuitry that initiates Reactor Building isolation is part

 ~          of the engineered safeguards protective system and is designed to meet Criteria 15 and 16. (Section 71.3.2)

Isolation valves ar.d valves which control other engineered safeguards equipnent have test provisions, and periodic manual application of test signals is used to verify functional operability. 1.h.23 CRITERION 23 In detemining the suitability of a facility for a proposed site the acceptance of the inherent and engineered safety afforded by the systems, materials and ccruponents, and the associated engineered safeguards built into the facility, vill depend on their demonstrated performance capability and reliability and the extent to which the operability of such systems, materials , canponents , and engineered safeguards can be tested and inspected during the life of the plant. Ansver: All engineered safeguards systems are designed so that a single failure of an active cmponent vill not prevent operation of that system or reduce the capability below that required to maintain a safe condition. Two independent Reactor Building cooling systems, each having full heat removal capacity, are used to prevent overpressurization. (Sections 6.2 and 6.3) O The makeup and purification, and decay heat removal systems have redundancy of equipnent to insure availabiltiy of capacity. (Section 6.1) Sane engineered safeguards systems have both a normal and an emergency function, thereby providing nestly continuous testing of operability. During normal operation, the standby and operating units vill be rotated into service on a scheduled basis. The answer to Criterion 17 gives more detail regarding redundancy, testing, and nomal and emergency operation of engineered safeguards. Engineered se feguards equipment piping, which is not fully protected against missile damage, utilizes dual lines to preclude loss of the protective function as a result of the secondary failure. (Section 6) 1.2.2h CRITERION 2h All fuel storage and vaste handling systemn must be contained if necessary to prevent the accidental release of radioactivity in amounts which could affect the health and safety of the public. Answer: - The spent fuel storage pool is located within the fuel handling and storage area of the Auxiliary Building. The liquid vaste holdup tanks and the gaseous vaste storage and disposal equipment are located within a sepdrate ares of the same sbuilding. Both of these areas provide confinement capability in the event of an accidental release of radioactive materials, and both - are ventilated with discharges to the station vent. Analysis has demonstrated that the accidental release of the maximura activity content of a gaseous

      , vaste storage tank vill not cause doses in excess of the limits set forth in 10'CFR 100. (Section 11.1.2 5 3)
   .i  o...

1 23 .' ' .

                                                                                   'b000037 1

Radioactive liquii effluent leakage into the Nuclear Services Cooling Water System vi* :e detemined by monitors on the cooling vater discharge lines. Any ac .autal leakage frcs liquid vaste storage tanks vill be , ccBected in a sunt ud transferred to other tanks to prevent releases to the environment. (Sec' ion 11.1.2.h) 1.k.25 .dITERION 25 The fuel handling and storage facilities must be designed to prevent criticality and to maintain adequate shielding and cooling for spent fuel under all anticipated nomal and abnomal conditions, and credible accident conditions. Variables upon which health and safety of the public depend must be monitored. Answer: All refueling operations vill be carried out with the fuel under borated water to provide cooling and shielding for the fuel assemblies. Visual control of all fuel handling operations vill exist at all times except during fuel transfer fra the Reactor Building. Spent fuel is trans-ferred under water through a opent fuel transfer tube to the spent fuel storage pool. Storage space is provided in the pool for 1-1/3 cores and the spent fuel shipping cask. Additional undervater storage for large internal emponents is provided inside the Reactor Building . refueling canal. To avoid accidental draining of the spent fuel storage pool, there are no penetrations that would pemit the pool to be drained below a safe , level. The fuel transfer tubes between the spent fuel storage pool , and the refueling canal are provided with gate valves and gasketed closure plates to prevent leakage. Water depth in the spent fuel storage pool provides sufficient shielding for nomal occupancy by operating personnel. The spent fuel storage pool cooling system contains provisions to maintain water cleanliness, temperature, and water level. A 21 in. by 21 in lattice arrangement is used for the spent fuel storage racks to insure fuel assembly suberiticality, l.k.26 CRITERION 26 Where favorable enviremental conditions can be expected to require limitations upon the release of operational radioactive effluents to the environment, appropriate holdup capacity must be provided for retention of gaseous, liquid, or solid effluents. Answer: The radioactive vaste disposal system vill collect, segregate, process , and dispose of radioactive solids, liquids , and gases in such a manner as to insure, empliance with 10 CFR 20.

                                                                    . e i                  '

t000 338 1-2h t . A I

() The storage capacity for liquid and gaseous vastes vill provide up to 30 day hold up time for radioactive decay. Solid vastes vill be pro-4 cessed in a batch manner for off-site disposal. Liquid and gaseous vastes released to the environ =ent vill be =enitored and discharged with suitable dilution to assure tolerable activity levels on the site and at the site boundary. Gaseous vastes vill be stored, allowed to decay, and then released at a controlled rate to ensure lov ac-tivity levels and lio?id vastes will be diluted by mixing with the cooling tower blevdown. Wastes vill be sa= pled to establish release rates consistent with environmental conditiens. l.h.27 CRITERION 27 The plant must be provided with systems capable of monitoring the release of radioactivity under accident conditions. Answer: Radioactive gaseous effluents which =ay be released into enclosed areas are () collected by the ventilation systems and discharged to the statien vent through charcoal and particulate absolute filter. Permanently installed ares 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 supple-mental surveys, if necessary. Radioactive liquid effluent leakage into the .ervice water systems will be de-termined by monitors on the cooling water discharge lines. These monitors are used for normal operational protection as well as for accident conditions. The effluent frem the liquid vaste disposal system is sampled prior to discharge and the release to the environment is monitored to insure compliance with 10 CTR 20. 15 RESEARCH AND DEVELOPET REQUIRD(ENTS The research and development programs that have been initiated to establish final design or to demonstrate the capability of the design for future opera-tion at a higher power level are summarized as follows: 1 5.1 ONCE-THROUGH STEAM GENERATOR TEST Steady state and load changing ope rations will be performed to demonstrate the ability of the unit to follev the transients and the interaction of the con-trol system with the water level, steam pressure and flows. The test, equip-ment consists of one 37-tube, full-length unit and one 19-tube, fpil-length 3 unit. Previously,. a full-length, 7-tube unit was tested. The tubes were fab-O^' ricated in accordance with the production techniques anticipated fo/ the full- . si:ed unit. j' # l-25 (Revised 12-8-67) [)(l()()

l I-The latter portion of the program includes tests to determine the natural frequency of the tubes and other parts in the steem generator. This vill be acccuplished by artifically induced vibrations from all external source, and the, tubes vill be examined for evidence of tube-te-tube contact and wear at support points. 1 5.2 CONTROL ROD DRIVE LINE SST The test assembly for this program is a full-sized fuel assembly with associated control rod and control rod guide, adjacent internals , and control rod drive. The unit is being tested under conditions of temperature, pressure, flow, and water chemistry specified for the full-sized reactor installation. This program vill embrace a prototype phase in which the unit vill be subjected to misalignment, varying flow, and temperature. The second phase of this program is one of life-testing where the unit vill be continuously cycled to cover the number of feet of travel and the number of trips ant'icipated for its life in the reactor. Both phases of .the program vill confim the operability of the drive line in nomal and misaligned conditions, confirm the rod drop times and load carrying characteristics of the control rods and fuel assemblies , and detemine the wear characteristics of all the drive line ecmponents. 153 SELF-POWERED DETECTOR TESTS The test units for this pregram are the self-powered detectors described in T.3 3. These units have been tested in T.he B&W Test Reactor at conditions of temperature and neutron flu:: anticipated in a central station reactor. These units are currently being tested in the Big Rock Point Nuclear Power Plant where they are exposed to temperature, neutron flux, and flow for conditions approximating those in the Three Mile Island Nuclear Station. The resulta j of these pregrams will provide a detector system with predictable characteristics of performance and longevity under incere conditions. 1.5.k THERMAL AND HYDRAULIC FROGRAMS 1 l B&W is conducting a continuous research and development program for heat transfer and fluid flow investigations applicable to the design of Three Mile Island Nuclear Station. Two important aspects of this program are: ~

a. Reactor Vessel Flow Distribution and Pressure Dron Tests A 1/6-scale model of the vessel and internals is 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.

b. Fuel Assembly Heat Transfer and Fluid Flev Test Critical heat flux data have been obtained on a single channel tubular and annular test sections with unifon and non-uniform heat fluxes on the multiple rod fuel assemblies with uniform i heat fluxes. These data have been obtained for a rar4e of pressure, temperature, and mass velocities enecmpassing the reactor design conditions. This work is being extended to include multiple red fuel assemblies with non-unifom axial heat generation. Additional mixing, flow distribution, and ,

pressure drop data vill be taken on models of various reactor flow cells and en a full-scale fuel assembly. .

     ~                                                                                                                                       ,

1-26

  .g   i                                                                                                                                       0000 340

i 1.6 IDENTIFICATION OF AGENTS AND CONTRACTORS Met-Ed vill be responsibP for the design, purchasing, construction, and operation of Unit 1, Three Mile L and Nuclear Station. This practice has been sucessfully followed for all of the Company's major generating facilities nov in service or planned. . The GPU System Nuclear Group has been organized to mobilize most effectively ' the capabilities and experience of GPU System's personnel for vork on the various nuclear projects in the GPU System. The organization established for this prcject is shown on Figure 1-13. The Project Manager is responsible for the coordination of all matters pertaining to site selection and development, des'ign engineering, preparation of reports required by various agencies and regulatory bodies, equipment purchasing, and construction associated with this project. Major equipment vill be procured by Met-Ed Purchasing Department on the basis of specifications prepared by the Architect-Engineer. The Met-Ed Production Department has the responsibility for training, pre-operational testing, and maintenance of the stations The Project Manager vill cooperate with the Superintendent of Production to assure a fully effective program. , The Nuclear Review Board, composed of engineers and managers experience with n other nuclear projects in the General Public Utilities system, as well as

   \j         certain specialists available to this project, is intended to provide a review of all major design decisions from the standpoint of safety and reliability.

The firm of Pickard, Love & Associates has been retained as Nuclear Consultants to assist in the preparation of reports and studies, to serve as design review specialists and to furnish guidance in nuclear related matters associated with the securing of the required permits for the project. Gilbert Associates, Inc. has been retained by Met-Ed as the Architect-Engineer for this project and vill assemble the necessary information for all required site studies and plot plans. In addition, they vill furnish all plant layouts and system arrangements and vill draw up specifications for major equipment and systems and cooperate with Met-Ed in the evaluation of all bids. ( United Engineers and Ccustructors, Inc. has been retained as Construction 1 Manager to supervise and coordinate the construction of the plant. UE&C will verk in cooperation with GAI to establish a schedule for the orderly procurement and delivery of equipment and material to the job site. The firm of Sheppard T. Powell and Associates has been retained as consultants on vater chemistry. Personnel are available, as required, to serve as design l review specialists. l l Met-Ed has contracted with Babcock & Wilcox to design, manufacture, and deliver to the site the ecmplete nuclear steam supply system. In addition S&W vill supply ecmpetent technic'al and professional supervision of erection, of initial V fuel leading, and of testing and initial start-up of the complete nuclear steam supply system. B&W will also ccoperate with Met-Ed in the training

c. J 0000 d'41

 -  a 2.^'i,.**

1-27 (Revised 12-8-67)

and licensing of Met-Ed operating personnel prior to and during the startup and initial operating period. MPR Associates, Incorporated, has been retained as consultants for Quality h Control and vill conduct a ec=prehensive quality control engineering effort covering (a) the nuclear steam supply system, (b) the containment system, (c) the fuel handling system and (d) the radioactive vaste disposal system. The firm of Weston Geophysical Research, Inc. was engaged to perform a seis-micity study and to develop response spectra for the site. - Mr. Richard J. - -- - Holt administered this work with Rev. D. Linghan, S. J. directing the seismi-city analysis and Professor Robert V. Whitman of Massachusetts Institute of Technology developing the response spectra. Dr. Marvin E. Kaufftan, of Franklin and Marshall College, was engaged to research the regional structural, geologic, and tectonic aspects of the site. Nuclear Utilitica Ser'vice Corp. have performed meterological computations with consultation from Dr. B. Davidson of New York University.

1.7 CONCLUSION

S The personnel assembled to design, construct, and operate the Three Mile Island Nuclear Station are competent. It is their combined intention to make this a conservative design and one which can be operated to produce electric power safely and economically. Toward this end -

a. The site has been examined and found to be suitable for the ~ '~ ~~ '~

nuclear station. The station at this site is compatible with ~ surrounding population and land uses, present and expected. Site characteristics of meterology, hydrology, geology, and seismology are favorable.

b. The reactor system chosen is a practical design of proven type, and its expected performance vill not require fuel exposures or energy-release rates higher than those presently

 ~

proved achievable using materials now available. Its shutdown margin and performance characteristics are comparable to those ~~ ~~ ~ ~ used in existing reactors. Before it commences commercial operation, the reactor system vill be thoroughly tested to confirm the desirable features designed into it, and that it vill perform as extected with full safety margins.

c. The reactor vill be installed in an enclosure both modern and conservative in design, which vill be able to contain and control all materials, vapors, or energies which could con-ceivably be released as a result of an accident. Supplementing the enclosure capability vill be engineered safeguards which will reduce to a minimum the consequences of an accident and insure that the dynamic conditions existing after an accident are kept within design parameters.

d. The station vaste and emergency systems vill be designed to release only effluents permitted by the AEC Regulations. Where ,

practicable, liquid and gaseous vastes vill be treated so that

       ,   ,.         the effluents contain a minimum of radioactivity and significantly
       ,e      .*     less than that allowed by applicable regulation.

1-28 (Revised 12-8-67) MOD -]H I

O e. A training prcCram is planned which vill adequately prepare operating personnel so that they will be qualified to test, start-up, and operate the nuclear unit. Experience gained in the design, construction, and operation of the Saxton Nuclear Reactor and New Jersey Central Power and Light's Oyster Creek Nuclear Station vill be of considerable value. Jersey Central is one of Met-Ed's sister ecmpanies within the parent organization of General Public Utilities (GPU). In consideration of the above circumstances and plans , it is concluded that the proposed Three Mile , Island Nuclear Station can be designed, constructed, and operated in a safe manner; that the proposed design vill provide adequate protection to the public fra any sequence of events resulting in disablement of equipnent from causes , natural or mechanical; and that Metropolitan Edison Capany and their consultants are qualified to design, construct, start, operate, and maintain this proposed nuclear generating unit in accordance with all applicable laws and regulations and in a manner satisfactory to the Atmic Energy Camission, to the public interests and to itself. O l l O 1-a 0000 043

Table 1-1 Engineered Safega -ds_ l Function Total Eq;irment Installed High Pressure Injection 3 pumps (makeup) 1 storage tank l3 Core Flooding System 2 tanks Lov Pressure Injection 2 pumps ( decay heat remaval) 2 heat exchangers l6 Reactor Building Spray 2 pumps System 1 sodium thiosulfate tank Reactor Building Cooling 2 pumps 6 System 3 emergency cooling units O l O -

                                                                       *',0000 044 1-30 (Revised 1-8-68)

1 I

     **                                                                                                                                      Tabla 1-2
    ..,                                                                                                                            Om1=rt a<a of Deslam farameters (per stattom unit teals unless muted) three Mlle latasd              ocomme ab clear htton
        -                                           Item                                                                Buclear Station                    th811 or 2        Turkey Ibiet Ib. 3 er 4          Intina 70 tat Ih. 2 1        Rrtrealle ased Thermal Dro1ga Parameters

- . finted Beat htpiat, Mit 2,452 2,49 2,097 2,758 listed Beat htput, m/br 8,369 a 106 8,3(9 m gf 7,g$7 , gf 9,gg 3 , gf tensism overpuuer, $ 14 14 12 12 System Fressure (mm an1), pela 2,200 2,200 2,2$o 2,2$o System Freehre (alata,a steady state), pela 2,150 2,150 2,220 2,220 ftsver Distributton Factors goat camerated la Fuel sat Claattag, $ 97 3 97 3 97.4 3 97.4 (austent) 1.85 1.85 1 73 1.75 maclear) 3.15 3.15 3 12 3 12 But aa-1 Factors Fg (auc. and mech.) 3 24 3 24 3 25 3 25 Die sentie at stated Conditione 2.27 fW-3) 2.27 w.3) 1.85 (W-3) 1.01 (W-3) 1.(o i BAW-1(a) 1.60 BaW.168) staansen LHB ltatto at Destga overla,uer 1 73 fW-3) 1 73 W-3) 130 (W-3) 1 30 (W-3) 1.38 f BAW-iti8) 1.3S (AAW-168) tuulant F1w

                                                                     ,4tal F1m Rate, Ib/br                        131 3 a 10,6                      3 33,3 , go6           300,6 , 396                   136.2 a 106 Effectiv- Flow Itate for Beat transfer, Ib   220 9 a lo"                       120 9 a 106            915 a lob                     124.1 a 100 ttfactive Flow Area for seat Transfer, ft    47 75                             47 75                  39.0                          48.4 Average Velocity Along Fuel Rods, ft/sec     15.70                             15 70                  13 9                          16 1 Average penas Velocity, ab/hr-ft2            2.53 a 10 6                       2 53 a If              2 35 a 10 6                  2.4 a 106 cwleat Tempratore, F Ikalan1 Inlet                                555                               555                    546.5                         543 maatam inlet due to tantrum atattua arror eat Demsbans                         557                               557                    550 5                         547 7

g average plu sa Venel Average anae as cure 47 8 49 3 47 8 49 3 59 53 57 p Average la Core $79 7 579 7 577 572 7 Average la Veeeet 578.9 578 9 574 570 Ikataal outlet of not a,manel 644.4 644.4 647 643 Average Falm hafficient, Stu/br-ft 2-F 5,coC 5,ouo 5,500 5.94 Average Fila Temprature httference, F 31 31 30 30 Ilsat Treamfer at 100$ fuwer Active neat Transfer Sarface area, ft2 48,578 48,578 42 460 52,200 , Average Beat Flum, m/Lt-ft2 167,620 167,6 m Ild,2m 175,600 hism Beat Flum, m/hr-f te 543,000 9 3,000 533,M M ,800 Average Themel output, hv/ft 5.4 56 53 51 hasm thermal output, kw/ft 17 5 17 5 17 3 18.5 msma clan asface waperature at iluminal Frusure, F 6A 6A 657 659 Fuel central Temperature, 7 hism at loo $ Iwer 4,160 6,160  %,o70 6,150 unalem at 114% overpower 4,400 4,600 b,270 4,250 themal output, kv/ft at hima overpuwer 19 9 19 9 19.4 20 7 2 Cbre Beechanical Destga Phremeters Fuel Assentelles Dastga Qia cea CEA can RCE caalese ' RCC caaleas aud Fitch, sa. o.558 o.558 o.563 c.563 e

.~. ) .

O O , C , . w O O, i e

V m d

     .A.

D', Table 1-2 (Cbat 'd ) Dree Itale talaat Ocunee helear Statica item lhaclear Statica that 1 and 2 Turkey haat au. 3 or 4 laitaa h1at b. 2 overall Dimenssoas, sa. 8.522 a 8 522 8.522 a 8.522 8.426 a 8.k26 8.k26 a 8.426 het weight (as Do2), ab 201,520 201,520 179,mo 215,320 h tal weicht, Ib 283,200 283,200 226,200 273,410 maber of crida per Asaeably 8 0 8 8 het Ada thater 36,816 36,816 32,028 39,372 Outalde Diameter, ta. 0.420 0.b20 0.422 0.422 Diametral Cap, Sa. 0.006 0.006 0.0065 0.0065 Clad n icknema, la. 0.026 0.026 0.0243 0.0243 Clad m terial Eircaloy-b Eirceloy-4 Eircaloy Eircaloy F.sel Fv11ets h terial UOg statered Do2 alatered tog sistered Dog afatered tenasty, % of theoretscal 95 95 94-93 94-9 D1.e.eter, sa. o.362 0 362 o.3%9 o.3fe)9 1.enct.h. in. o.8 c.8 o.600 0.600 Control sa,1 As.emblies (Can) Neutron Abaorber  % On.M$ la-M% 4  % CJ-1% !a-80% Ag $ Cd.1% !a-80% Ag  % Q11% !a-80% 4 Cladding h terial 304 SS - cold worked 304 SS - cold worked 304 SS - cold worked 304 SS - cold worked Clad Thickness, ta. 0.018 0.018 0.019 0.019 Ikeber of Assemblien 69 69 41 53 R&mber of Control Ikale per Assunbly 16 16 20 20 Core SLnecture I g Core Imrrel ID/CD, in. 1%f/150 147/150 133 5/137.25 148 5/152.5 ta neraal mield ID/co, In. 155/159 155/159 141.0/1475 158.5/164.0 to 3 Preltatmory sk.cicar nestca Data structural caractersatics N 1 wear,ht (as 10 2 ), in 201,520 a01,520 179,000 215,3Lo Clad Weicht, Ib 43,000  %),000 35,600 %3f 785

                             %re Diameter, in. (equivalent)                   128.9                              128.9                    119 5                   134 Core Nescht, la. (active fuel)                    1%                                 1%                       1%                      1%

e heflector Wickness asm1 Composition 50 TLp(waterplussteel),ta. 12 12 10 10 O btton (water plus steel), ta. 12 10 10

                                                                                                                .32 h                             Side (water plus steel), ta.

M2 0/U (unit cell - cold) 18 18 15 15 on 2.97 2.97 3 48 3 48 M Ika. sher of Ibel Assemb11em 177 177 157 193 A' hel Ada/Phel Assembly 208 208 204 204 g perferiaance Characteristica p taadind Technique 3 region 3 regica 3 radica ) regica 1 Fuel Discharge braup, ledD/MIU Ch Averace First cycle 12.460 82 14,000 12,000 k4 Equilibrium Core Average Feed Ihrictaments, w/o U.235 28,200 28,60 h.I 200 27,000 27,000 v Begica 1 2.25 EEE" 2.28 Region 2 2.2} 2.64 2.47 2.43 2 35 meston 3 2.90 2.77 2 73 2.68 r+111brium 2.94 3 09 -- 2.92 Control Qaaracteristica Q ~ Effective N1ttp11 cation (beginalag of life) No. 1 Rb. 2 ~~ O Cold, Ib her, Cleam 1 .302 EDY E25Y 1.275 1.275 1 Not, aim her, Clean 1.247 1.258 1.201 1.223 1.225 g Bot, Dated her, ze ami as agu11sbrim 1 .1SC 1.lti7 1.119 1.170 1.170 _ O O .aa

                           %p
                                                                          .e Table 1 2 (Cunt

• d )

                                                                                ,                                                                                        three K11e Islassi              Ocence Lclear Statten Itaa                                                       hclear Stattua                      that 1 c.r 2          hikey Fuist Ib. 3 or %         !sstina h ir.t Ib. 2 coatic.1 aua Aase=.611ee h tersal                                      55 CJ-15% In-6 5 A4               5> C4-15$ In.ai$ A4         55 01 15$ In-80$ A4          S$ 5 15% In-00$ As hat.or of Asseah!!es                          69                                69                          41                           53 Ed.er of Amr Rude per CSA                     16                                16                          20                          20 total b1 mrth (Akg  T /, $                     10.0                             30.0                        7.0                         70                             1 brea Concentratloaa h S.ut Reactor Duwe With Rode IaeerteA (cleaa) cola / bot pian                  1290/1000                        12')0/1150                  2300/2500                    3400/3500 Ibrum Wrth (let),             pps              1/10u                            1/100                       1/1p                        1/1p brom Wrth (cola), $             pp             1/75                             1 /73                       1/100                       1/120                       __

E.1astic Omrecteristice bierator haperature Coefficient, F el.0 a 10~% to -5.0 a 10 +1.0 a 10'% to .3.0 a 10*% +1.0 a 10-% to -3 0 a 10-% + 1.0 a 10' to .3 0 a 10*% bleratc.r Pressure Coefficient, pei -1.0 a 10-0 to e3 0 a 10 -1.0 a 10*0 to e5 0 a 10-0 -1.0 a 10-6 g ,3,o a go-6 1.0 a 10"0 to 93 0 a 10*0 b ierator Vota Coefficient, $ veld +1.0 a 10*b to .3 0 a 10*3 +1.0 a 10-4 to -3 0 a 10-3 +0 5 a 10-3 o -2.0 a 10*3 t +1.0 a 10*3 to .3 0 a 10*3 Doppler coefficient, F 1.1 a 10"I to -1 7 a 10-5 1.1 a 10-5 to .1 7 a 10-5 1.0 a 10-5 to .2.0 a 30-5 1.0 a 10-5 to -2.0 a 10-3

                                                                                                                %   Principal Desica 34rameters of the henctor Coolant System System acat oatput, ears                        2,468                             2,468                       2.097                       2,758 7

tg system acat o.tput, stu/hr Operati c Pressure, peig 8,k23 a ad 2.185 8,k23 a 10 0 2,185 7,156 a 10 6 2 23 9,412 a 106 2,235 (4 seactor Inlet bsserature, F 555 555 .5b.5) 5%) Beactor oatlet Temperature, F 603 603 600.6 9b.0 Ikeber of Ecope 2 2 3 testsa Pressure, peig 2,500 2,500 2,485 2,485 Deatge Teerperature, F 650 6$0 650 650 Ertreatatic Test Pressure (cola), pela 3,125 3.125 3.110 3,110 Coolant Volime, tactuatag pressurtser, ft I 11,800 11,800 9,800 12,209 tbtal peactor Flow, sp 352,000 352,000 266,400 358,800 h (6 5 Pea-tcr cbolaat ftrete Coa mequiremente h n Beacter Vessel Steam Gstnerator A38 III, Case A ADE III, Class A A3 E III, Clase & A3 E III Casa A g) b be Slas ASE III, Qass A A3E III, Casa A A3 E III, C asa A A3E III Class A fa EcIl SLie A38 III, Case A ASE III, Qaae A ASE III, Qaas C A38 III, Casa C Prcasuriser ASE III, Class A ASE III, Qaae A AS E III, Class A A38 III, Case A h g Pressuriser Belief Tank Pressuriser Safety valves ASE III, Casa C ASE III ASE  !!!, AS E III Casa C A38 III, Qaae C ASE III ASE III, Qaae ASE III C p Reactor Coolant Piping seactor Coolant Puzy Castag ASA a31.1 ASE III, Case A ASA 331.1 A3 E III, C asa A ASA 331.1 ASA 331.1 D 6 Princip1 Deatgm Pareneters of the Beactor Vessel neaterial SA-)D2, Limaa 3, ela4 with SA-302, CreJe B, clat with SA-302, crede 3, clad with SA-302, Crede 3, alsA with Type 3% austanttle $3 type 30b austealtas SS type 304 emetaattle 83 type 304 austealtte 83 0 *. O-O O O

                                                  .c=

N O m O 9

t _ -g b

                           ~

Shle 1-2 (Oost'd ) 3ree title Island 0;onee ILclear Station g 12, clear Stattom tatt 1 or 2 Ariey hint ib. 3 or b Indina hast Ib. 2 Deessa Pressure, pata 2,500 2, W 2,4S) 2,kSi Destga temperature, F 630 630 f)O 6)0 Operettag Pressure, pela 2,10) 2,18) 2,23) 2,23) Inside Diameter of Sell, ta. til 171 155.5 173 Outside Diameter Across Mozales, ta. 249 243 240/233-3/8 24) Overall Insteht or Vessel ama closure seed, rt-ta. 41-8 5/8 41-8 3/8 43-0 42 k Mintana Clad Dicknese, ta. 1/8 1/8 3/32 }/32 7 Prtact:s t Dest;n Puremeters of the steen Generstars haber er thits 2 2 3  % Type Vertical, once-through Vertical, once-through Vertical, U-tukse with tate- Vertical, U-Lestee with late-with latelrml super- with later,ral super- gral maisture seprator. gret actsture separator, 1.6ater. I. ester. hbe senterial Iaconel Iacome! Income! ta.manal smell tenterial Carbon steel carbon steel carbon steel Carbon steel hbe side Destca Pressure, pets 2,500 2,500 2,485 2,485 bbe Side Destca haperature, F 650 650 650 650 g hbe Side Destsa Flow, Ib/br 65 66 a 106 65.66 a 1M 33.53 a ad 34.05 s ad g  ! Riel! Side Deste,n Pressure, pets 1,0$0 1,050 1,(45 1 c85 w Sell StJe Destga beperature, F 600 60A 600 600 C" Operettag Pressure, hbe Side, Iksm8 mal, pets 2,185 2,185 2,23% 2,2 35 operating Pressure, 3.elt GAde, lensimm, teig 910 910 1,005 1,005 lemaamm at>1sture at 0,ttet at Full Imed, 5 35 F an,perheat 35 F auperheat 1/4 1/4 Iqrdrostatic ut pressure (tube side-colu), pets 3.125 3,125 3,110 3,110 8 Prlactsal Desita hrometers of the Ilmactor Omalent FLaBe shaaber of thite 4 4 3  % Type Vertical, stacle stage Vertical, slagte stage Vertical, stag 3e stage. Vertical, single stage. Radial riou with botte Andial flow with bottom suct1on and boriscatal suct8on and borisontal etecharge, discharge. Destga Pressure, pstg 2,500 2,500 2,L85 2,685 Destga h aperature, F 650 650 650 650 08*ratind Pressure, Ikastaal, pata 2,185 2,185 2,235 2,235 shetton h aperature, F 555 }5_5 546.5 543 Destga capacity, cims 88,000 as,oco 88,800 83,700 Destga htal Develot*4 Seed, rt 370 310 256 272 Rydrostatte het Pressure (cold), pats 3,125 3,125 3.110 3,110 hter Type A-C Indactice, single A-C !adanction, staste 4-C Iaiuction, eingle A-C !aluettoa, single speed speed speed speed htor datina (austaal), by 9.000 9,000 5,500 6,000 9 rrt'ncssal Dessan parameters or the

                               ,a,                                   ,

peactor Coolant Ftpl u lemterial Omrtion steel clad with SS Carbon steel clad with SS dansteattle 30 Austealtss 33 loot tg (ID). ta. 36 36 29 29 C' Cold tag (ID), to. 28 28 27-1/2 27-1/2 O C3 O

                           -s=

CO

   *7 hble 1-2 (Cont'd)

• Three Mile Zalaat Oconee hclear Station Itas helear Stattaa thalt 1 and 2 hrkey Ebist h- 3 or 4 ledian hint b. 2 Between Ibugp and Steen Generator (ID),1s. 26 26 31 31 10 Ileactor ht14tru % stem Parameters
• Type Steel-lined, prestresset, Steel-lined, prostressed, Steel-18aed, prestressed. Steel-11and, retaforced post-tenelomed concrete, post-teasicaed concrete, poet-teassomed concrete, concrete, vertteal cylla-vertical cyltater with wertical cyltater with vertical cylleter with der with flat bottom ant flat bottom anal shallow flat bottom ami shallow flat. bottom amt shalluer basispherical dome.

domed roof. A-t roof. damed roof. Deelga hrometere Inside Diameter, ft 1p 11 6 116 135 Belebt, ft 206 } 107 177 212 Pree volume, ft3 2,000,0u0 3,900,000 1,550,000 2,610,000 Reference Incident Pressure, pais 55 $9 56 47 fleference lecident Energy (Eg ), Stu 306,700,000 306,700,000 272,000,000 M5,P90,000 Ehergy Required to Produce Incident Pressure (Eg), Ste 3)$,200,000 348.s06.000 300,000,000 349,8fb,000 a.etos si/Eg 0.915 0.s97 0.907 0.871 setto: (Eg-Eg)/E3 0.09) 0.8845 0.103 0.166 chacrete tfisekasse, ft Vertical Wall 3-1/2 3-3/4 }-1/2 $ 1/2 Ihme 3 3- 8 / 4 3 4-1/2 psactor Ihallding taak Prevention fash-taght penetrations Leak-tight penetraticas Leak-ttsht penetraticas coatt-ely preneurtsed asst > Mitigation as.1 continuume steel ami contianaoue steel ami coattauuue steel double penetratione, liner. Automatic isola. liner. Autumatic leola- 11aer. Automatic tecla- lineer weld channels and tion where required. tion where required. Es- tion where required. accese op sings. Isola. ( gg baust frum penetration tion valt a seal water roans to stattoa vent. esoteen automatScally See-Un lates piping, 4ere re. quiree. Oaattauoue leak rate monStoring of eca-Q,a talmesat eat pressurised

                                                                                                                                                                                                                      -e .

F Gaseous Effluent Purge Liecharas vent above top 04scharge went mee top Through particulate filters vest discharge fra top of 2 of iteactor b11 ding of Neector unalJse.g amt moatters. hit of containment (*150' above (-200 ft above grase) g 2Og is ano ,, geode) the amin enhaust system. grade). A. 11 Engtaeered Safecamsde id h g Safety Rajectica % sten Iks. of Bidh sent Piampe 3 5 3 3 m ab. of Imr Eead 3%mpe 2 v Os Beector bildla g nanergency Coolers. 3 2 2 lb ab. of thnite 3 3 3 5 Air Flotr Cap *y. rach, at Accident Cumittion, cfin 54,000 54,cuo 80,000 65,(no Core Ficodtag % stem - me. of hake 2 2 3 4 html volume, ft3 2,$00 2,$00 3,600 4,400 PL,etace14ent Filters None taside Reestor flualJ-

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