ML19319D678
| ML19319D678 | |
| Person / Time | |
|---|---|
| Site: | Crystal River, 05000303  |
| Issue date: | 08/10/1967 |
| From: | FLORIDA POWER CORP. |
| To: | |
| References | |
| NUDOCS 8003240650 | |
| Download: ML19319D678 (55) | |
Text
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f -TABLE OF CONTENTS Section Pg 1 INTRODUCTION AND SUWJLRY l-1
1.1 INTRODUCTION
1 1.2 DESIGN HIGHLIGHTS 1-2 1.2.1 SITE CHARACTERISTICS 1-2 1.2.2 POWER LEVEL 1-2 1.2.3 PEAK SPECIFIC POWER LEVEL 1-2 1.2.h Rl' ACTOR BUILDING l-2 1.2 5 ENGINEERED SAFEGUARDS 'l-3 1.2.6 ELECTRICAL SYSTD4S AND DERGENCY POWER ' l-3 1.2 7 ONCE-THROUGH STEAM GENERATORS 1-k 1.3 TABULAR CHARACTERISTICS l-h 1.h PRINCIPAL DESIGN CRITERIA y 1-7 1.4.1 CRITERION 1 1-7 1.h.2 CRITERION 2 1-9 1.4.3 CRITERION 3 1-10 1.h.h CRITERION h 1-10 1.h.5 CRITERION 5 1-11 1.k.6 CRITERION 6 1-11 1.h.7 CRITERION 7 1-12 1.4.8 CRITERION 8 1-13 1.h.9 CRITERION 9 l-13 1.k.10 CRITERION 10 1-lh 1.h.11 CRITERION 11 1-15' l.h.12 CRITERION 12 1-16 h 1.k.13 CRITERION 13 1-16 00000024 1-1
TABLE OF CONTENTS Section Page 1.4.1h CRITERION 1h 1-17 1.4.15 CRITERION 15 1-17 1.4.16 CRITERION 16 1-18 1.h.17 CRITERION 17 1-19 1 1.18 CRITERION 18 1-21 1.h.19 CRITERION 19 1-21 1.h.20 CRITERION 20 1-22 1.h.21 CRITERION 21 1-22 1.4.22 CRITERION 22 2-23 1.h.23 CRITERION 23 1-23 1.h.2h CRITERION 2h 1-24 O 1 6 25 ca rza1on 25 1-2h 1.h.26 CRITERION 26 1-25 1.h.27 CRITERION 27 1-26 1.5 RESEARCH AND DEVELOPMENT REQUIREMENTS 1-26 1.5.1 ONCE-THROUGH STEAM GENERATOR TEST 1-26 1.5.2 CONTROL ROD DRIVE LINE TEST 1-26 1.5.3 SELF-POWERED DETECTOR TESTS 1-27 1.5.h THERMAL AND HYDRAULIC PROGRAMS 1-27 1.6 IDENTIFICATION OF AGENTS AND CONTRACTORS 1-27
1.7 CONCLUSION
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O LIST OF TABLES (At Rear of Section) Table No_. Title M l-1 Engineered Safeguards 1-30 1-2 Comparison of Design Parameters 1-31 O i O ao000026 l 1-111 l l
}[ LIST OF FIGURES (At Rear of Section) )
Figure No. l Title 1-1 Service Area 1-2 Equipment List 1-3 General Arrangement Floor Plan Elevs. 95', 91' & 80' 1-4 General Arrangement Floor Plan Elevs. 119' & 122' 1-5 General Arrangement Floor Plan Elevs. 126'-5" & 143' 1-6 General Arrangement Floor Plan Elevs. Ih6' & 160' 1-7 General Arrangement Floor Plan Elevs. 178' & Roof Plans 1-8 General Cross Section A-A & B-B l-9 General Cross Section C-C 1-10 General Cross Sectiou D-D 1-11 Plot Plan ( - 1-12 Nuclear Project Management Organizational Chart 1-13 Power Department Organizational Chart 1-14 Mechanical Engineering Department Organizational Chart b o
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1 INTRODUCTION AND
SUMMARY
1.1 INTRODUCTION
This Preliminary Safety Analysis Report is submitted in support of Florida Power Corporation's application for a construction permit and facility license for the one unit nuclear addition to the Crystal River Plant loca- 5 ted on the Gulf of Mexico in Citrus County, Florida, and shall be designated . Unit 3. The plant location is shown on Florida Power Corporation's service 5 Area Map, Figure 1-1. The generating unit will operate initially at core power levels up to 5 2L52 Wt which c:rrespends to a gross electrical output of about 855 We. All physics and core thermal hydraulics information in this report is based upon the reference core design of.2452 MWt. It is expected that the nuclear supply system vill be capable of an ultimate output of 2560 MWt, (including 16 MWt contribution frca reactor coolant pumps), corresponding to a gross electrical capability of about 885 MWe. All power conversion systems will be designed to accccmodate the altimate unit output. Site parameters, prin-cipal structures, engineered safeguards, and certain hypothetical accidents are evaluated for the expected ultimate core output of 2544 MWt. The nuclear steam supply system is a pressurized water reactor type similar to systems operating or under construction. It uses chemical shim and control rods for reactivity control and generates steam with a small amount of superheat in once-through steam generators. The nuclear steam supply system and the first four fuel cores will be supplied by The Babcock & Wilcox Company. Construction of Unit 3 is schedulid for completion in time for fuel loading to begin by December 1,1971, for earliest commercial operation by April 1,1972, and latest co=mercial operation by October 1,1972. To meet this schedule, construction of Unit 3 is to begin by June 1,1968 and is to 5 be completed by March 1,1972. The general sa rangement of major equipment and structures, including the Reactor, Auxiliary, and Turbine Buildings, is shown on Figt.res 1-2 through 1-10. 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 completely responsive to that guide, to the proposed AEC design criteria, and to all known pertinent questions asked of other applicants up until the time of this writing.
As the plant design progresses from conceptual design to final detailed design, the plant description and analyses will be subject to change and refinement. This report presents descriptive material and analyses of a
" reference-design." Amy significant changes to the criteria or designs which affect safety vill be promptly brought to the attention of the AEC by revised insert pages, and additional information vill be submitted by suit-able supplements.
'O 1-1 (Revised k-8-68)
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Florida Power Corporation (FPC) is fully responsible for the co=plete safety and adequacy of the plant. Aid in the design, construction, testing, and start-up of the unit vill be supplied principally by Gilbert Associates, Inc. (GAI), and The Babcock & Wilcox Co=pany (B&W). Assistance shall als, g be rendered by other consultants and suppliers as may be required. The technical qualifications of FPC, B&W, and GAI are outlined in Appendix 1A. 1.2 DESIGN HIGHLIGHTS 1.2.1 SITE CHARACTERISTICS The k,738 acre site is characterized by a khoo foot minimum exclusion radius; isolation from nearby population centers; sound foundation for structures; 7 an abundant supply of cooling water; an ample supply of emergency power; and favorable conditions of hydrology, geology, seismology, and meteorology. 1.2.2 POWER LEVEL Initially licensed povar for the reactor core is proposed at 2,k52 MWt, and 5 core perfor=ance analyses in this report are based on this initial po rer level. Operating confir=ation of reactor core parameters is expected to support an ultimate core power level of 2,5hk MWt, and the unit vill be 5 designed to operate at this output. The analyses of accidents that could release fission products to the environ =ent have been evaluated on the basis of 2,5kk 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 855 MWe at initially licensed power and 885 MWe 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 co= parable with other reactors of this size and therefore does not represent an extrapolation of technology. This comparison may be seen in the infornation presented in Table 1-2 which is discussed in detail in Section 1.3. 1.2.h REACTOR BUILDING The leek tight structure required to contain the Maxi =u= Hypothetical Accident (MHA) is the Reactor Building. The prestressed, post-tensioned concrete Reactor Building is of essentia11y' the sa=e design as the containment buildings for the Turkey Point Plant (Docket Nos. 50-250 and 251), the Oconee Station (Docket Nos . 50-269, 270, and 5 287), and the Three Mile Island Nuclear Station (Docket No. 50-289). Several of the engineered safeguards are similar to those plants and Crystal River Plant Unit 3 and present neither unco ==en solutions to engineering problems ncr significant extrapolations in technology. i 1-2 (Revised 7-15-69)00000029 0
i 1.2 5- ENGINEERED SAFEGUARDS' Engineered safeguards are employed to reduce the potential radiation dose to the general public from the Maximum Hypothetical Accident to less than the guide- ' line values of 10 CFR 100. This is acccuplished by automatic isolation of all reactor building penetrations that are not required for limiting the con-i sequences of the accident, thus eliminating potential leakage paths. Long-tera 4 potential releases following the accident are reduced by rapidly decreasing the , reactor building pressure to near atmospheric within 24 hr, thereby reducing 1 the driving potential for fission product escape. In addition, the engineered safeguards will prevent core meltdown should the worst postulated loss-of-coolant accident occur. This is accomplished by injec-l' tion core flooding systems of large capacity, parts of which are continuously operated for normal purposes and are therefore immediately available for emer-gency duty. These systems, coupled with the thermal, hydraulic, and blowdown I characteristics of this reactor, will reliably prevent metal-water reactions l5 and any core melting (or disfiguration of the core into a geometry to prevent further cooling). The engineered safeguards equipment of the nuclear unit, along with the normal 5 operating modes, are as follows: l 4
, a. High pressure injection - normal 37 operates as part of the makeup and l purification system.
4
- b. Core flooding system,
- c. Low-pressure injection - normally operates for shutdown cooling as
! part of the decay heat removal system.
- d. Reactor building spray system -rormally shut down. '
- e. Reactor building cooling system - normally operating except for emergency cooling water supply.
- f. Reactor building isolation system - operates on test or accident sig-nal.
Table 1-1 lists equipment supplied for the engineered safeguards. 1.2.6 ELECTRICAL SYSTDIS AND EMERGENCY POWER Crystal River Plant Unit 3 will have the following sources of electric power: 5
- a. The generator will feed its own auxiliaries. Upon a trip separating the substation from the transmission system, the turbine-generator is designed to stay in operation upon a load dump from full load down to auxiliary load.
0GQQ0 Othrw e T O , . -. SM 1-3 (Revised k-8-68)
- b. Four 230 kv transmiasion lines--two to Curlev and two to Central Florida.
- c. u ther or both of Units 1 ale 2 at the same site. O
- d. Two quick-starting diesel-generator units connected to safeguards l5 busses. These are presently estimated to be rated 2850 kv each. l1 Within the plant there vill be multiple redundant busses and ties supplying power to loads, instruments, and controls. The engineered safeguards are sup-plied from two separate safeguard power busses, each of which can be supplied l5 from any of the four principal sources of power.
The sources of power and associated elect-ical equipment vill insure safe fune-tioning of the plant and its engineered safegua is. 1.2.7 ONCE-THROUGH STEAM GENERATORS The steam generators are of a new design based on extensive research, devel-opment, and experimental work on boiling heat transfer performed by B&W over the pnst 10 years. Each generator is a vertical shell-and-tube, counterflow heat exchanger with reactcr coolant on the tube side and steam on the shell side. Feed rater is pumped into the generator, heated to saturation by direct mixing with stear, converted to steam and superheated in a single pass through the generator. The basic design parameters, such as feedvater heating, boil-Ing length, superheat length, and performance characteristics, have been con-fir =ed by testing of a full-length 7-tuba unit and a 37-tube unit. Tests are continuing to provide additional data in these design areas for the 37-tube test unit. In addition, testing vill continue with one, 19-tube full-length unit. l1 g 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. j 1.3 TABULAR CHARACTERISTICS l l Table 1-2 is a cceparative list of important design and operating characteris-tics of the Crystal River Plant Unit 3, Three Mile Island nuclear Station l (Metropolitan Edison Co=pany), Oconee Huclear Station Units 1 and 2 (Duke l5 l Power Ccepany), and Turkey Point Units 3 and h (Florida Power and Light Com-l pany). The design and operating parameters of the Metropolitan Edison, Oconee, and Turkey Point stations are close to those of the Crystal River facility. ! The Oconee and Metropolitan Edison units each have the same rated core power as the Crystal River facility and are near-duplicates in other respects. The data in Table 1-2 represent information presented in available plant descrip-tior.s, and Safety Analysis Reports submitted for licensing. The design of each of these plants is based on information developed from op-i eration of commercial and prototype pressurized water reactors over a number i of years. The Crystal River design is based on this existing power reactor technology and has not been extended beyond the boundaries of known informa tion or operating experience. t
~
1h (Revised k-8-6
1 i l I The similarities and differences of the features of the reactor plants listed ' in Table 1-2 are discussed in the following paragraphs. In each case, the item number used refers to the item numbers used in the table. Item 1. Hydraulic and Thermal Design Parameters The parameters listed in this section are the same for the first three columns and are similar to the Turkey Point units. The differences in power level are reflected chiefly in the total heat output, core size (fuel loading), 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 frcan Nucleate Boiling Ratio (DNBR) and the maximum ratio of peak-to-average. total heat input per fuel rod (F Ah nue.) are representative of a more conservative design for the B&W reactors than for the other reactor presented. These ccuparisons are discussed in detail in 3.2.3.2. Item 2. Core Mechanical Design Parameters The dimensions, materials, and technology for each of these reactors are sim-lar. (Note the same design basis for Crystal River, Oconee, and the Three Mile Island units.) This uniformity is again due to optimization of the op-erating 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 between the first three columns (B&W units) and the Turkey Point units. The increased number of control rods in the B&W reactors provides for maneuverability and flexibility of operation. The B&W reactors utilize a canned fuel assembly which provides structural integrity and protection of the fuel rods against damage during fuel handling operations. Item 3. Preliminary Nuclear Design Data Since these reactors have essentially the same core geometrical configuration, the fuel loading differs by an amount proportional to the physical size of the reactor core. Note that the design data for the Crystal River and Three Mile Island units and the Oconee units, are the same except for the first-cycle burnup, feed enrichments, control characteristics, total rod worth, and boron concentrations. These differences reflect the greater first-cycle burnup of Three Mile Island and Crystal River over Oconee. Oconee Unit 1 has a Crst-cycle fuel sbaring program with Unit 2, which requires a lover first-cycle enrichment for Unit 2. 5 The basis of the 2.97 (H2 0/U) is the ratio of the water molecules to the atoms of U-metal in the fuel assembly envelope. This value of the water to U-metal volume ratic, consistent with the entry for Turkey Point Units 3 or 4, is 3 53. Each core has a three-region fuel loading, but differs in the fuel burn-up ratio that is to be used. 000000 0000002 gJ ;; . 1-5 (Revised 4-8-68)
The Crystal River and Three Idle Island reactor designs offer about 2 5 per cent greater reactivity control in the control rods over Turkey Point. 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. Principal Design Parameters of the Reactor Coolant System Most of the features in this section are directly related to material proper-ties and the amount of heat produced in the reactor core. Note that the B&W units are identical. The parameters are scaled in proportion to the power of the reactor. The major difference is the number of coolant loops required to remove the heat produced. For the B&W units only two loops are s equired, since once-through steam gener-ators are used instead of the U-tubes-in-shell design. The greater cooling capacity of these steam generators permits a reduction in the number of cooling loops for an equivalent amount of heat removed. Item 5 Reactor Coolant System - Code Requirements The B&W units are identical. Code requirements for the shell side of the steam generator conform to the ASME III Class A specification. This is considered to be a contribution to the safety of the vessel, since it enhances the integrity because of the more stringent ASME III Class A design, material, and quality control requirements. Ite: 6. Principal Design Parameters of the Rea.: tor Vessel 2he B&W units are identical. These vessel designs are characterized by a thinner O 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 Principal Design reatures of the Steam Generators The steam generators in the B&W units are the same. They are basically different from the Turkey Point units since they are once-through design and incorporate an integral superheat section. Item 8. Principal Design Parameters of the Reactor Coolant Pumps l The B&W designs are the sa=e. In each specific tabular parameter the relative nunber or size is in proportion to the total amount of heat removed from the core. The B&W reactor pu=ps have higher head and horsepower requirements than the Turkey Point units have for approximately the same flow because of the in-creased flow losses of the once-through steam generators and the use of only two reactor coolant loops. Ites 9 Principal Design Parameters of the Reactor Coolant Piping l The B&W unit piping designs are the same. They utilize carbon steel clad with stainless steel. 00L00033 0
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Item 10. Reactor Building Parameters
'w/ All reactor buildings are basically of the same design and construction. The differences are physical dimensions, amount of concrete shielding needed, and design incident pressures, which are a direct result of plant layout, engi-neered safeguards, system capacities, and site location. The reactor building design and shielding offer satisfactory protection to the surrounding popula-tion in case of an accident and during normal operation of the generating units.
Item 11. Engineered Safeguards Engineered safeguards are generally similar, but Oconee includes a penetration room ventilation system for each unit. The Oconee and Turkey Point units do not use sodium thiosulfate injection in the reactor building spray for iodine removal because of the differences in design and site characteristics. If it 3 is deemed necessary to use a filtering system for iodine removal, space vill be available in the present reactor building design for installation of such a system. l.h PRINCIPAL DESIGN CRITERIA The Crystal River Plant Units 3 and h vill be designed to meet the 27 General Design Criteria for Nuclear Power Plant Construction Permits (l) proposed by the Atomic Energy Commission and the proposed 70 Design Criteria (2) as inter-preted by Florida Power Corporation. The principal safety features that meet each criterion are summarized herein for the 27 General Design Criteria and in Supplement No. 2, Informal Question h for the propor ed 70 criteria. In the discussion of each 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 tu the mitigation of their consequences must ce designed, fabri-cated, and erected to: (a) Quality standards that reflect the importance of the safety function to be performed. It should be recognized, in this respect, that de-sign codes commonly used for nonnuclear applications may not be ade-quate. (b) Performance standards that vill enable the facility to withstand, without loss of the' capability to protect the public, the additional forces imposed by the nost severe earthquakes, flooding conditions, vinds, ice, and other natural phenomena anticipated at the proposed site. Answer: The integrity of systems, structures, and couponents essential to accident pre-vention and to mitigation of accident consequences has been considered in the
' design evaluations. These systems, structures, and components are:
- 1. Fuel assemblies.
- 2. Reactor vessel internals.
(1)The criteria as proposed by the AEC in its press relesise H-252 of November ( 22, 1965. ,. i I2)The criteria as propo' sed by the AEC in its press release K-172 of July 3 10, 1967. 1-7 (Revised 3- *
- ilff/,);,m. -1; i) -
- s. -
l l l 3 Reactor coolant system.
- 4. Reactor instrumentation, controls, and protection systems.
5. 6. Engineered safeguards. Radioactive materials handling systems. lll
- 7. Reactor building.
- 8. Electric power sources.
(a) Quality Standards The fuel assemblies are designed to maintain their integrity when subjected to the mechanical, hydraulic, and thermal stresses resulting from anticipated oper-ating conditions during their design life. The design is based on technology which has proved successful in existing nuclear power plancs and is substanti-ated by test data. The fuel assemblies vill be manufactured to high quality rtandards and subjected to a series of rigorous tests during fabrication. (Section 3.2.h.2) Components and piping in the reactor coolant system are designed for a pressure of 2,500 psig at a temperature of 650 F. The nominal operating conditions of 2,185 psig and 579 F allow 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 h l.5) Piping and Valves - ASA B31.1 (Piping Code) including nuclear cases. Pump 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 Yessel - 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 4.k) l The fast neutron exposure (3 x 1019 nyt) of the reactor vessel results in a maximum NDTT of not more than 260 F at the end of Plant service life. Operating procedures for the Plant will be compatible with these tem-perature limitations. (Section k.1.4) The instrumentation, reactor control system, and protection systems will be fabricated using high quality components and workmanship. All compo-nents will be selected on the basis of reliability, durability under service conditions, and proven application. In the absence of applicable industry standards and codes regulating quality in this type of equipment, quality standards are established based on extensive experience, study and testing. (Section 7) Engineered safeguards, i.e. , emergency injection, core flooding system, the Reactor Building spray system, the Reactor Building isolation system, and the Reactor Building emergency cooling units, are designed and will be fabricated to high quality standards and tested to insure proper operabil- j O l-8 010000035
- I ity. All piping in these systems is designed to the applicable ASA Code for pressure piping. (Section 6)
Radioactive material handling systems are designed and will 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 Reac+or Building is designed and vill be constructed in accordance with applicable sections of appropriate ACI and ASTM codes and specifica-tions as well as criteria described in Section 5 E The unit will have redundant normal sources of electric power which will supply the essential electrical load requirements as follows: (a) its own auxiliary transformer fed directly from the main generator, (b) the 230 kv transmission lines, (c) either or both of the existing units through either startup transformer. Four redundant battery-backed busses will be provided for vital instruments. tion and contrcl. Electrical equipment vill be purchased and tested to stringent requirements for reliability and quality, including appropriate NEMA, USASI, and IEEE electrical standards. (Section 8) (b) Performaneg Standards All equipment and structures having a safety function vill be designed, con-structed, operated, and maintained without loss of captbility to protect the public under all environmental conditions anticipated at the site. 1.4.2 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 signi-ficant amounts of fission products from the facility. Answer: Energency injection coolant is provided te maintain the core sufficiently cov-ered to prevent core melting for the complete range of postulated reactor cool-ant system rupture sizes up to the maximum size of a 36 in. ID pipe. In the process of cooling the core, the metal-vater reaction is limited to an insigni-ficant amount. (Section 14) l Makeup pumps with a capacity of approximately 1,000 gpm vill inject water into the reactor coolant system. This system is primarily effective early in the accident while the reactor coolant system pressure is above 100 psig. (Section 6) A core flooding system is supplied to cool the core at intemediate to low pres-sures. Tuis system consists of two independent core flooding tanks, each of which is connected to a different reactor vessel injection nozzle. o 000000M l-9 (Revised k-8-68)
-__em-, __ - _ . ,. _ _ _ - . _ __ _ _ _ __ _ _.
Decay heat pumps provide 9,000 gpm of water to cool the core when it is partially or totally uncovered, and the reactor coolant pressure has dropped below 100 psig. (Section 6) 1.k.3 CRITERION 3 Protection must be provided against possibilities for damage of the safeguard-ing features of the facility by missiles generated through equipment failures inside the containment. Answer: Protective valls and slabs, local missile shielding, or restraining devices will be provided to protect the Reactor Building liner plate and engineered safe-guards within the Reactor Building against damage from missiles generated by equipment failures. The concrete enclosing the reactor coolant system seJves as radiation shielding and as an effective barrier against missiles. Local missile barriers vill be provided for control rod drives and where required be-cause of oper.ings in the concrete shield enclosing the reactor coolant syetem. (Sections k and 5.1.2.7.1) For those parts of the safeguards susceptible to missile damage, redundant equipment is provided to insure required operation. (Section 6) 1.4.4 CRITERION k The reactor must be designed to acccamodate, without fuel failure or primary system damage, deviations from steady state norm that might be cecasioned by abnor=al yet anticipated transient events such as tripping of the turbine-generator and loss of power to the reactor recirculation system pumps. j Answer: The reactor is designed with a margin above normal operating conditions i to accommodate anticipated abnormal deviations from the steady state operation. ~his margin allows for deviations of temperature, pressure, flow, reactor power, and ro ' tor-turbine power mismatch. The reactor is operated at a constant average ecolant temperature above 15 per cent power and has a negative power coefficient to dampen the effects of power transients. The reactor control system vill maintain the reactor opera-l ting parameters within preset limits, and the reactor protection system l vill shut down the reactor if normal operating limits are exceeded by i preset amounts. (Sections 7.1 and 14) The nuclear unit is shut down automatically by the reactor protection system if a co=plete 1 css of electrical power occurs. Upon loss of ex-l ternal system electrical load, a reactor power reduction occurs, and the reactor continues to generate Plant 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 atmosphere, thereby preventing excessive reactor coolant system pressures. Accordingly, these transients will not produce fuel i or reactor coolant system damage. (Sections 7.1 and 14.1.2.8) 1-10 00000M7
l i I
.; The reactor. coolant pumps are provided with sufficient-inertia to main-- ,
tain adequate flow to t:revent fuel damage if power to all pumps is lost. The criterion for core protection following loss-of-coolant flow is to maintain a Departure from Nucleate Boiling Ratio (DNBR) equal to or greater than that at the design overpower level for initial power condi-tions up to and including the maximum operating power level of 107.5 per cent power. Natural circulation coolant flow will provide adequate core cooling after the pump energy has been dissipated. (Section 14.1.2.6 and Figure 9-10) 1.k.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 readily suppressed. Answer: The ability of the reactor control and protection system to control the oscillations resulting from variation of coolant temperature within the control system dead band and from spatial xenon oscillations has been analyzed. Variations in average coolant temperature provide negative feedback and enhance reactor stability during that portion of core life in which the moderator temperature coefficient is negative. When the coefficient is positive, rod notion vill compensate for the positive feedback. The maximum pove change rate resulting from temperature os-cillations within the contrO. system dead band has been calculated to be less than 1 per cent / minute. Since each unit has been designed to fol-low ramp load changes of 10 per cent per minute, this is well within the capability of the control system. Control flexibility with rdspect to xenon transients is provided by the combination of control rods, incore instrumentation, and out-of-core in-strumentation. Within control rod limits, transient xenon related to load changes is controlled by the automatic control system. Axial, radial, or azimuthal neutron flux changes will be detected by both out-of-core and incere instrumentation. Individual or groups of control rods can be posi-l tioned to suppress and/or correct flux changes. (Section 3.2.2.2.3) i ! 1.h.6 CRITERION 6 Clad fuel must be designed to accommodate throughout its design lifetime all normal and abnormal modes of anticipated reactor operation, including the design overpower condition, without experiencing significant cladding failures. Unclad or vented fuels must be designed with the similar ob-jective of providing control over fission products. For unclad and vent-ed solid fpe,lg vormal and abnormal modes of anticipated reactor operation mustbeachievg Qt exceeding design release rates of fission products from the fuel over re, lifetime. Answer: p -Fuelcladintegrityisinsuredunderallnormalandabnormalmodes[of anticipated operstion by avoiding clad overstressing and overheating, 1_11 00000038
i The evaluation of clad stresses includes the effects of internal and external pressures, temperature gradients and changes, clad-fuel inter- h actions, vibrations, and earthquake effects. The free-standing clad de-sign prevents collapse at the end volume region of the fuel rod and provides sufficient radial and end void volume to accommodate clad-fuel interactions and internal gas pressures. (Section 3.2.h.2) Clad overheating is prevented by satisfying the following core thermal and hydraulic criteria: (Section 3.2.3.1.1) (a) At the design overpower no fuel melting vill occur. l (b) A 99 per cent confidence exists that at least 99 5 per cent I of the fuel rods in the core vill be in no jeopardy of experi-encing a DNB during continuous operation at the design over-power of llh per cent. 1.4.7 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 or electrical control system malfunction could cause a reactivity transient capable of damaging the primary system or causing significant fuel failure. Answer: Reactivity control vill be accomplished by movement of control rods and h by changes in soluble poison (boron) concentration in the reactor cool-ant. 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 malfuncticas. Approach to criticality and low power operation vill be by manual rod withdrawal. The remaining rods (or rod groups ) vill be interlocked to permit withdrawal on automatic control only after the rod groups used ! for approach to criticality and low power operation have been fully with-drawn. Rods used for automatic control vill be arranged in four groups and interlocked to prevent simultaneous withdrawal of more than two groups. That is, sinultaneous withdrawal of two automatic groups vill be permitted over approximately the first 25 per cent of the second rod group stroke and the last 25 per cent of the first rod group stroke. The maximum reactivity insertion rate associated with simultaneous with-drawal of a regulating twelve rod group is 5.8 x 10-5 ss k/k/sec. Assum-ing a single electrical failure occurs that invalidates the interlock and pennits the 25 control rods on automatic control to be withdrawn simultaneously, a maximum reactivity insertion rate of 2.3 x 10-4 Zsk/k/ sec. could result. Reactivity transients of this magnitude have been analyzed, and the resultant power transients will not produce reactor coolant system'or fuel failure. (Section 1k.1.2.3) 00000039 4 1-12
A reduction in the reactor coolant soluble poison concentration vill require operator initiation, and vill be prohibited by interlocks until the control rods are in an acceptable pattern for dilution. A second safety feature b^ will physically limit the maximum rate at which dilution water can be added . to the Jystem. A third safety measure vill consist of a relay which will limit the total time of dilution. The maximum reactivity insertion rates from moderator dilution vill be 7.0 x 10-6 a k/k/sec. These rates are not sufficient to produce damage to either the fuel or reactor coolant system. l.h.8 CRITERION 8 Reactivity shutdown capability must be provided to make and hold the core suberitical from any credible operating condition with any one control element at its position of highest reactivity. Answer: The reactor is designed with the capability of providing a shutdown margin of at least 1 per cent A k/k with the single most reactive control rod fully withdrawn at any point in core life with the reactor at a hot zero power condition. The minimum hot shutdown margin of 2.1 per cent Ak/k occurs at the end of life. l3 Reactor suberitical margin is maintained during cooldown by changes in soluble poison concent11 tion. The rate of reactivity compensation from boron addition is greater than the reactivity enange associated with the maximum allowable reactor cooldown rate of 100 F per hour. Thus, suberiticality is assured during ecoldown with the most reactive control rod totally unavailable. O V 1.h.9 CRITERION 9 Backup reactivity shutdown capability must be provided that is independent of normal reactivity control provisions. This system must have the capability to shut down the reactor from any opere, ting condition. Answer: Soluble poison addition vill provide an independent backup to the control rods for reactivity shutdown. Poison addition vill be accomplished using the makeup and purification system. There are three makeup pumps to insure flow availability under all credible operating conditions. These pumps take suction from the borated water storage tank, which contains water with 2,270 ppm boren, or frem the makeup tank. In the latter case, a solution containing 8,750 ppm boron is supplied to the makeup tank from a mixing tank. Two transfer pu=ps 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 from any operating condition. The following table demonstrates the capability of shutdown without control rods for two modes of makeup and pur-l ification system operation. i .ijjf. O ' " ' 00000040 ! 1-13 (Revised 3-1-68)
,.... , v; . v . ,
l . l
Soluble Poison Shutdown Capability Negative Reactivity Time to Shut Down From 1005 Feed Insertion Rate, g Concen- Feed (% ek/k)hainute Rated Power t9 Hot Zero Power Condition L1), minutes W tration, Flow Rate, ppm boron gpm BOL EOL BOL EOL 8.750 20 0.0179 0.0217 67 106 1 h,121 70I2) 0.0224 0.0353 5h 65 3,196 lh0I3) 0.0278 0.05k3 h3 h2 1.h.10 CRITERION 10 H;at removal systems must be provided which are capable of accommodating core d:;; cay heat under all anticipated abnormal and credible accident conditions, such as isolation from the main condenser and complete or partial loss of pri-mary coolant from the reactor. Answer: Rzactor decay heat vill be removed through the steam generators until the reactor 3 coolant system is cooled to 250 F. Steam generated by decay heat vill supply the main and emergency feedvater pump turbines and can also be vented to atmosphere or bypassed to the condenser. The steam generators are supplied feedvater by(l) l1 flow the rate r.ain steam-driven for decay feedvater heat renoval,I2) pt ps, whichemergency by a steam-driven can be operated feed pump, at a orreduced (3) by a motor driven emergency feed pump energized from the emergency diesel-generator system. The main feedvater pumps supply water contained in the feedvater train and the condensate storage tank to the steam generators. The emergency feed pumps take suction from the condenser hotwell and the condensate storage l3 tank. These sources provide a minimum of 200,000 gal of water storage per nuclear unit which is sufficient for decay heat removal for about one day l1 after reactor shutdevn with the condenser isolated. The condenser is nomally 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 1h.1.2.8) (1) Reactivity balance on Doppler and moderator equal to 1.2% ak/k for BOL 1 and 2.3% 4k/k for EOL. ( Makeup to makeup tank at 20 gpm of 8,750 ppm boron from boric acid mix tank plus 50 gpm at 2,270 ppm boron from stcrage tanks. (3) Makeup to makeup tank at 20 gpm of 8,750 ppm boron from boric acid mix tank plus 120 gpm at 2,270 ppm boron from storage tanks. 00000041 ) 1-14 (Revised 3-1-68) 1
- ~. - -
Under conditions of ' complete or partial l'oss-of-coolant from the reactor,
~
decay heat will be removed from the core by coolant supplied by the emer-gency injection coolant systems. The source of injection water will be the borated water storage tank. When this source is exhausted, the low pressure injection pumps vill take suction from the Reactor Building sump. The return flov is cooled and pumped to the reactor vessel to continue core cooling. This system contains redundancy of equipment to insure availability of flow when required. If complete loss of external electric power occurs, on-site sources supply sufficient electric power for all engineered safeguards and cooling water systems. (Section ik.2.2.3) 1.4.11 CRITERION 11 Components of the primary coolant and containment systems must be designed and operated so that no substantial pressure or thermal stress vill be imposed on the structural materials unless the temperatures are well above the nil-ductility temperatures. For ferritic materials of the coolant envelope and the contain-ment, minimum temperatures are NDT + 60 F and NDT + 30 F, respectively. Answer: The reactor vessel plate material opposite the core is purchased to a specified NDTT of 10 F cr less and is tested to verify conformity to specified require-ments. (Section b.2 5) The end-of-unit-life-NDTT value of the reactor vessel opposite the core vill ( be not mere than 260 F. Unit operating procedures will be established to limit the operating pressure to 20 per cent of the design pressure when the reactor coolant system temperature is belov NDTT plus 60 F throughout unit life. Surveillance specimens of the reactor vessel shell section material vill be installed between the core and inside vall of the vessel shell to monitor the NDTT of the vessel material during operating lifetime. (Section h.l.k) The reactor vessel material is protected from excessive radiation damage by coolant water annuli between the core and the reactor vessel. The thickness of these annuli limits the total fast flux greater than 1 Mev incident on the reactor vessel vall to an nyt value of 3 x 10 19 in ho years at an 80 per cent unit capacity factor. The thermal shield contributes to a further reduction in vessel material radiation damage. (Section h.1.h) The Reactor Building liner is enclosed within the Reactor Building and thus vill not be exposed to the temperature extremes of the environs. The Reactor Building ambient temperature during unit operation vill be between 100 and 110 F which is expected to be well above the NDT temperature +30 F for the liner plate. The liner plate is completely enclosed by the thick concrete valls, slab, and roof of the Reactor Building, and thus will not be sub.iect to sudden variations due,to changes in external temperatures. In addition, the bottom liner plue ik protected by a minimum thickness of 21 in. of con-crete. The pene'tratiorij vill be made of material which exhibits, by test, a transition tempeNture at least 30 F below the minimum service metal # temperature. For the purpose of determining minimum service metal temper-ature, the lowest one day mean temperature vill be assumed as 25 F. 0048@& ^
% o. L 1-15
1.k.12 CRITERION 12 Capability for control rod insertion under abnormal conditions must be provided. Answer: Control rods will provide the normal means for changing reactivity to shut down to a hot suberiticsl 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 sig-nals 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 from a hot to a cold zero power condition. The principal safety criteria for the control rod drives are: (a) No single failure in the drive shall result in the loss of safety function. (b) Trip action shall not require power, and no single failure or chain of failures shall prevent trip action to more than one mechanism. (c) The trip command shall override all other commands. Trip action shall be nonreversible. The reactor vessel, reactor vessel supports, reactor vessel internals, fuel assemblies, control rods, and the control rod drive are all de-signed to resist, without loss of function, the effects of seismic load-ings established by the seismological analysis of the site. h The control rod is never withdrawn completely from the fuel assembly. The guide structure is oriented with respect to the fuel assembly by a common grid sT,ructure which maintains full stroke control rod guidance into the fuel assembly. The drive line is designed and will be tested to be fully operable under conditions of the maximum misalignment speci-fied. (Section 3.3.3.h.1) 1.4.13 CRITERION 13 1 ! The reactor facility must be provided with a control room from which all actions can be controlled or monitored as necessary to maintain safe operational status of the plant at all times. The control room must be provided with adequate protection to permit occupancy under the con-ditions described in Criterion 17 below, and with the means to shut down the plant and maintain it in a safe condition if such accident were to be experienced. Answer: l The reactor unit will be controlled from a separate control panel located in the control room. The control room is designed to permit continuous occupancy following an accident. (Section 7.4.5)
)
O
,_u 00000043
' All controls and instrumeiitation require'd to monitor and operate' the reactor and electric power generating equipment vill be located within the control room. This includes indication of power level; process variables such as temperatures, pressures and flows, valve positions; and control rod position.
All engineered safeguards equipment will be controlled and monitored from the control room. The status of all dynamic equipment (pumps, valves, etc.) as well as pertinent pressures, temperatures, and flows will be displayed. The plant radiation ronitoring system will have instrumentation readouts displayed in tne control room. During MHA conditions the concrete Reactor Building and control room walls and roof provide adequate protection against direct radiation to control room personnel. Control room personnel on eight hour shifts during a 90 day period following the MHA would not receive an integrated direct dose in excess of 3 rem, from all sources of radiation which is approximately equal to the calendar quarter dose permitted in 10 CFR 20. The control room is provided with independent ventilation and filtration systems to control ingress of airborne contaminants escaping from the Reactor Building. (Section 9.7.2) The activity in the control room during any 90 day occupancy vill not exceed the limits specified in 10 CFR 20. , e The control room is also protected against the credible external missiles
\ more fully described under Section 5 \
1.h.1h CRITERION 14 Means must be included 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 room. The reactivity status of soluble poison vill be indicated by the position of the control rods. The soluble poison concentration 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 will receive results of laboratory analyses of the soluble poison concentration. (Section 7) 1.k.15 CRITERION 15 A reliable reactor protection system must be provided to automatically initiate appropriate action to prevent safety limits from being exceeded. Capability must be provided for testing functional operability of the system and for determining that no component or circuit failure has oc-
~
curred. For instruments and control systems in vital areas where the s potential consequences of failure require redundancy, the redundant
) channels must be independent and must be capable of being tested to de-
[V termine that they remain independent. Sufficient redundancy must,be 000U006 1-17
provided that failure or removal from service of a single component or i channel vill not inhibit necessary safety action when required. These 'g ' criteria should, where applicable, be satisfied by the instrumentation W i associated with containment closure and isolation systems, decay heat ' removal and core cooling systems, systems to prevent cold-slug accidents, and other vital systems, as well as the reactor nuclear and process l safety system. Answer: The reactor protection system is designed to provide the features specified in this criterion. A minimum of four sensors is 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. If a portion or all of an instrumentation channel is removed from service, the channel assumes a tripped condition. One channel in a tripped condition 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 from e 3-channel system described in Section 7. The power supply for each individual channel vill be from one of the 4 redundant battery-backed vital busses. (Section 8) The channels are normally energized, and loss of power to two busses causes a reactor l trip. (Section 7) Provisions will be included for testing the protection systems and/or components under administrative control on a periodic basis. Normal testing vill include the insertion of a simulated signal to dynamically check response and performance of each channel's components except de-tectors. Tests of each protection system channel vill insure a high confidence level of system operability. (Section 7.1.3.5) 1.h.16 CRITERION 16 The vital instrumentation systems of Criterion 15 must be designed so l that no credible combination of circumstances can interfere with the performance of a safety function when it is needed. In particular, the effect of influences common to redundant channels, which are intended to be independent, must not negate the rperability of a safety system. The effects of gross disconnection of the system, loss of energy (electric 00000045 # 1-18
p ( power, instrument air), an'd adverse environment (heat from loss of i'n-strument 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 Building ambient conditions ranging from h0 F to lho F vithout adverse effects in accuracy. Reactor Building temperature will be normally controlled in the range of 60 F to 110 F. The protection system instrumenta-tion, exclusive of the neutron detectors in the Reactor Building, vill with-stand the externe' pressure and temperature for the duration of a loss-of-coolant accident and still _a operable (but subject to several per cent in-accuracy). The out-of-core neutron detectors are designed for continuous operation in a temperature 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 protection 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 normally de-energized controls. They will be activated through redundant controls and power systems. (Section 7.1) es Manual reactor trip is designed so that failure of the automatic reactor trip circuitry vill not prohibit or negate the manual trip. The same is true with respect to manual operation of the engineered safeguards equipment. (Section 7.1) 1.h.17 CRITERION 17 The containment structure, including access openings and penetrations, must be designed 'and fabricated to accommodate or dissipate without failure the pressures and temperatures associated with the largest credible energy
? release including the effects of credible metal-water or other chemical reactions uninhibited by active quenching systems. If part of the primary l coolant system is outside the primary reactor containment, appropriate l safeguards must be provided for that part if necessary, to protect the health l 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, addi-tional containment, etc. , will depend on environmental and population con-ditions surrounding the site.
Answer: The Reactor Building, including access openings and penetrations, has a l 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, will not exceed 4 these values. () c' .
\ .i x_z -
000Mb 0 0 0. 'n l-19 . r-m wr- -_._w g .
.w--vy+- ---.wi-.wy we
The Reactor Building and engineered safeguards systems have been evaluated for various combinations of credible energy releases. The analysis accounts g for system energy, decay heat, metal-water reactions, and the burning of the W resultant hydrogen. The cooling capacity of either Reactor Building cooling system (see Criterion 18) is adequate to prevent overpressurization of the structure, and to return the Reactor Juilding to near atmospheric pressure within 24 hours. The details of this evaluation are discussed in Section 14.2.2.3.5. The use of injection systems for core flooding will limit the Reactor Building pressure to less than the design pressure. If a metal-water reaction is un-inhibited 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 pene-trate 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 vould require operator action to prevent escape of coolant external to the Reactor Building. This is a procedure that the operator would normally perform, i The makeup and purification system diverts a small zmount of reactor coolant i outside of the Reactor Building. This high pressure and high temperature coolant is cooled before it leaves the Raactor Building. Lines serving this function contain isolation valves that can be closed to prevent uncon-trolled 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 (Figure 9-5). Any leakage of reactor coolant g'
- through the letdown coolers vill be into this system ,rather than to the W I
environment. The intermediate cooling system is monitored to detect leakage of reactor coolant. Leakage of contaminated coolant from engineered safeguards equipment located external to the Reactor Building has been evaluated, and the resultant en-vironmental consequences are well below 10 CFR 100 limits at the site boundary, i and have been included in the total accidental dose calculations. l l The high pressure injection and decay heat removal systems have redundancy of equipment to insure availability of capacity. (Section 6.1) Certain engineered safeguards syste=s have both a normal and an emergency function, thereby providing nearly continuous testing of operability. For example, two makeup pumps are in continuous use for seal injection and makeup; the decay heat removal pumps are in use for decay heat removal during each shutdown; and one Reactor Building intermediate cooling system pump is in continuous use. During normal operati~ h the standby and operating units will be rotated c h into service on a ; scheduled basis. In cases where separate equipment is used solely for emergency conditions, such as the Reactor Building spray pumps, recirculating lines are provided, and instrumentation is in-stalled to provide means for conducting tests. The equipment is located to facilitate inspection during operation. (Section 6) 6 1_,, 00000047 1
' Electric motors, valves, and damper operators, which must function within the Reactor Building during accident conditions, will operate in a steam-air
, atmosphere at 281 F and 55 psig.
1.k.18 CRITERION 18 Provisions must be made for the removal of heat from within the containment structure as necessary to maintain the integrity of the structure under conditions described in Criterion 17 above. If engineered safeguards are needed to prevent containment vessel failure due to heat released under such conditions , at least two independent systems must be provided, pre-ferably of different principles. Backup equipment (e.g., water and power systems) to such engineered safeguards must also be redundant. Answer: Reactor Building cooling following the loss-of-coolant accident is provided by two independent systems: (1) the Reactor 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 condi-tions. The Reactor Building spray system supplies 2,000 gpm from the borated water storage tank into the Reactor Building. After the borated water storage tank is emptied, recirculation from the Reactor Building sump begins. This recir-culated water is cooled in heat exchangers by the nuclear services cooling Q D water system. The nuclear services cooling vater system is always in operation, and therefore has continuously indicated ave.ilability. Two sets of nozzles, located in the upper portion of the actor Building I 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 heat to the nuclear services cooling water system. Pumps and heat exchangers are redundant to insure availability. (Section 6.2) l Upon loss of external sources of electric power, one of the two diesel l5
- generators will permit operation of all required engineered safeguards equip-ment. -(Section 8.2.3) 1.k.19 CRITERION 19 The maximum integrated leakage from the containment structure under the condf-l tions described in Criterion 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 completion and installation of all penetrations, and the leakage rate measured j -over a stitable period to verify its conformance with required perfomance..
The plant must be designed for later tests at suitable pressures. ,
,*f'
,, G fO 00000m Mb\ 1-21 (Revised 4-8-68)
Answer: The Reactor Building leakage rate vill be determined at design pressure after completion and installation of all penetrations. The leak rate test will ' verify that the maximum integrated leakage does not exceed .25 per cent by weight of the contained air per 24 hours at the design pressure. (Section 5.1.2.2) The environmental hazards from the maximum hypothetical accident, assuming the above specified maximum integrated leakage from the Reactor Building, is within the guide line values of 10 CFR 100. (Section 14.2.2.4) A leak rate test at a suitable pressure using the same method as the initial leak rate test can be performed at any time during the operational life of the plant, provided the plant is not in operation and precautions are taken to l protect instruments and equipment from damage. All penetrations are desi6ned with double barriers and constructed so that leak rate tests can be performed on the individual penetrations at any time during the life of the plant. A program for periodical penetration testing during the operating life of the plant will be made to check the integrity of the penetrations and the operabil-ity of the pressurizing system. 1.4.20 CRITERION 20 All containment structure penetrations subject to failure such as resilient seals and expansion bellows must be designed and constructed so that leak-tightness can be demonstrated at design pressure at any time throughout operating life of the reactor. Answer: O All Reactor Building penetrations with resilient seals or expansion bellows vill be constructed so that they may be pressurized so that leak tests can be conducted at any time. (Sections 5.1.2.6.1 and 5.4) 1.k.21 CRITERION 21 l Sufficient normal and emergency sources of electrical power m'at be provided to assure a capability for prompt shutdown and continued maintenance of the reactor facility in a safe condition under all credible circumstances. Answer: The design of the electrical system for this nuclear plant vill provide sources of electric power as follows:
- a. Power to the nuclear plant from Unit 3 generator 5
- b. Frce remote units via 230 KV transmission lines
- c. From Units 1 or 2 via the 230 KV substation
- d. Two emergency diesel engine generators h
, 1-22 (Revised k-8-68)
'e #
Any one of' these power sources will have sufficient capacity to maintain the plant in a safe condition. 1.k.22 CRITERION 22 Valves and their associated apparatus that are essential to the containment function must 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 fuctional operability of these valves and asso-ciated equipment to determine that no failure has occurred and that leakage is within acceptable limits. Redundant valves and auxiliaries must be in-dependent. 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 loss-of-coolant accident. Reactor Building isolation occurs on a signal of approximately h psig or by manual actuation from the control room. The criterion for isolation valve requirements is: Leakage thrcugh all fluid penetrations ineluding those serving accident-consequence-limiting systems is to be minimized by a double barrier so that no single credible failure or malfunction of an active component can result in a loss of . Solation or intolerable leakage. The double barriers take the form or closed piping systems both inside and out-side the Reactor Building and various arrangements of isolation valves. (Section 5.2) 4 Fluid penetrations serving engineered safeguards systems also meet this criterion, but the actuators can be manually operated from the control room for test purposes. The control circuitry that initiates Reactor Building isolation is part of the engineered safeguards actuation system and is designed to meet Criteria 15 and 16. (Section T.1.3.2) Isolation valves and valves which control other engineered safeguards I equipment have test provisions, and periodic mant tl application of test signals is used to verify functional operabilie 1.h.23 CRITERION 23 In determining the suitability of a facility for a proposed site the accept-j ance of the inherent and engineered safety afforded by the systems, materials and components, and the associated engineered safeguards built into the ifahf1py g , vg1 depend on their demonstrated performance capability and rellAbility; add the extent to which the operability of such systems, materials, components, and engineered safeguards can be tested and inspected during the life of the plant. 00000#50 1-23 t
)
Answer: l All engineered safeguards cystems are designed so that a single failure of an active component will not prevent operation of that system or ' reduce the capability below that required to maintain a safe condition. Two independent Reactor Building coc1tpr systems, each having full heat removal capacity, are used to prevent overpressurization. (Section 6.2) The makeup and purification, and decay h(.ht removal systems have redundancy of equipment to insure availability of capacity. (Section 6.1) Certain engineered safeguards systems have both a normal and an emergency function, thereby providing nearly continuous testing of operability. During normal operation, the standby and operating units will be rotated into service on a scheduled basis. The answer to Criterion 17 gives more detail regarding redundancy, testing, and normal and emergency operation of engineered safeguards. Engineered safeguards 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.h.2h CRITERION 2h All fuel storage and vaste handling systems 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 separate area of the same building. Both of these areas provide confinement capa-bility in the event of an accidental release of radioactive materials, and both are ventilated with discharges to the plant vent. Analysis has demonstrated that the accidental release of the maximum activity content of a gaseous vaste storage tank will not cause doses in excess of the limits set forth in 10 CFR 100. (Section 11.1.2 5.3) Radioactive liquid effluent leakage into the Nuclear Services Cooling Water System vill be determined by monitors on the cooling water discharge lines. Any accidental leakage from liquid vaste storage tanks vill be collected in a sump and transferred to other tanks to prevent releases to the en-vironment. (Section 11.1.2.4) 1.h.25 CRITERION 25 The fuel handling and storage facilities must be designed to prevent crit-icality and to maintain adequate shielding and cooling for spent fuel under all anticipated normal and abnormal conditions, and credible accident con-ditions. Va-iables upon which health and safety of the public depend must be monitored. 0000005I > noogo.m O 1-2h
7S Answer: U 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 will exist at all times except during fuel transfer from the Reactor Building. Spent fuel is transferred under water through a spent fuel transfer tube to the spent fuel storage pool. Storage space is provided in the pool for a minimum of 350 fuel assemblies and the spent fuel shipping cask. Additional underwater storage for large internal components is provided inside the Reactor Building re-fueling canal. To avoid accidental draining of the spent fuel storage pool, there are no penetrations that would permit the pool to be drained below a safe level. The fuel transfer tubes between the spent fuel storage pool and the re-fueling 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 normal 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 uted for the spent fuel storage racks to insure fuel assembly sub-criticality. 1.h.26 CRITERION 26 () Where unfavorable environmental conditions can be expected to require limita-tions upon the release of operational radioactive effluents to the environ-ment, 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 compliance with 10 CFR 20. Solid vastes will be processed in a batch manner for off-site disposal. Liquid and gaseous vastes released to the environment vill be monitored 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 low activity levels and liquid vastes will be diluted by mixing with the condenser circulating vater discharge. Liquid vastes that have activity levels too high for direct discharge vill be held in storage tanks for decay and dilution or for evaporation. Ample holdup storage capacity for liquid vaste is installed and vill store all ef-fluent from operation. Wastes will be sampled to establish release rates consistent with environmental conditions. The gaseous vaste system vill store accumulated gas that is released during operation. The contents of the storage tanks vill be sampled, and a release rate eptablished consistent with the prevailing environmental conditions. ('T) Inline monitoring vill provide a continuous check on the release of activity.
.. . . v aw
.gd : vm t 1-25 00000052 .
9 e _, ,_ _ , _ , . _ _
1.k.27 CRITERION 27 The plant must be provided with systems capable of monitoring the release ,g of radioactivity under accident conditions. Answer: I Radioactive gaseous effluents which may be released into enclosed areas are collected by the ventilation systems and discharged to the plant vent through charcoal and HEPA filters. Permanently installed area de-tectors and the plant vent detector are used to monitor the discharge levels to the environment. In addition, portable moniters are available on site for supplemental surveys, if necessary. Radioactive liquid effluenf. leakage into the service water systems vill be determined by monitors on the cooling vater discharge lines. These monitors are used for normal operational protection as well as for accident conditions. The effluent from the liquid vaste disposal system is sampled prior to dis-charge, and the release to the environ =ent is monitored to insure compliance with 10 CFR 20. 1.5 RESEAP.CH AND DEVELOPMENT REQUIRIMENTS 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-TEROUGH STEAM GENERATOR TEST Steady state and lead-changing operations will be performed to demonstrate the ability of the unit to follow the transients and the interaction of the control system with the water level, steam pressure, and flows. The test equipment consists of one 37-tube, full-length unit and one 19-tube, full-length unit. Previously, a full-length, 7-tube unit was tested. The tubes l1 were fabricated in accordance with the production techniques anticipated for the full-sized unit. The latter portion of the progra= includes tests to determine the natural frequency of the tubes and other parts in the steam generator. This will be accomplished by artificially induced vibrations from an external source, and the tubes vill be examined for evidence of tube-to-tube contact and wear at support points. 1.5.2 CONTROL ROD DRIVE LINE TEST The test assembly for this program is a full-sized fuel assembly tith asso-ciated control rod and control rod guide, adjacent internals, and control rod drive. The unit is being tested under conditions of temperature, pressure, i flow, and water chemistry specified for the full-sized reactor installation. l This program will embrace a prototype phase in which the unit vill be sub-l jected to misalignment, varying flow, and temperature. The second phase j of this program is one of life-testing where the unit vill be continuously cycled to cover the nu=ber of feet of travel and the number of trips anti-l cipated for its life in the reactor. Both phases of the program will confira O l-26 (Revised 1-15-68) 00000M3
the operability of the drive line in normal and misaligned conditions, con-firm the rod drop times and lead-carrying characteristics of the actuator, indicate vibration and fretting wear characteristics of the control rods
)
and fuel assemblies, and determine the wear characteristics of all the l drive line components. I 1 5.3 SELF-POWERED DETECTOR TESTS l The test units for this program are the self-powered detectors described in 7.3.3. These units have been tested in the B&W Test Reactor at conditions of temperature and neutron flux anticipated in a central plant 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 Crystal River Plant. The results of these programs vill provide a detector system with predictable characteristics of performance and longevity under incere conditions. 1.5.h THERMAL AND HYDRAULIC PROGRAMS B&W is conducting a continuous research and development program for heat transfer and fluid flow investigations applicable to the design of the Crystal River Plant. Two important aspects of this program are:
- a. Reactor Vessel Flow Distribution and Pressure Drop Tests A 1/6-scale model of the vessel and internals is under test to l1 measure the flov 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 Flow Test Critical heat flux data have been obtained on single channel tubular and annular test sections with uniform and nonuniform heat fluxes on the multiple rod fuel assemblies with uniform heat fluxes. These data have been obtained for a range of pressure, temperature, and mass velocities encompassing the reactor design conditions. This work is being extended to include multiple rod fuel assemblies with nonuniform axial heat generation. Additional mixing, flow distribution, and pressure drop data vill be taken on models of various reactor flow cells and on a full-scale fuel assembly.
1.6 IDENTIFICATION OF AGENTS AND CONTRACTORS
- Florida Power Corporation vill be responsible for the design, purchasing, construction, and operation of units 3 and 4 nuclear addition to the Crystal River Plant. This practice has been successfully followed for all of the Company's major generating facilities now in service or planned.
The organization established for this project is shown on Figure 1-12. The Nuclear Project Manager (Chief Mechanical Engineer) is responsible for j the coordination of all matters pertaining to site selection and develop- , l ment, design engineering, preparation of reports required by various agencies O o nogyoogg a . 4t i
- (,}
u 1-27 (Revised 1-15-68)
~ _ _ _ _ _ _ _ _ _ .- _ _ _ _ _ _ . .__,_.__ _ _ . . .
and regulc. tory bodies, cquipm:nt purchasing, ccnstruction, sind start-up associated with this project. Major equipment will be procured by Florida Power Corporation Purchasing Department on the basis of specificati,ons prepared by the Architect-Engineer (Gilbert Associates, Inc.). The Mechanical Engineering and the Production Dep:trtment operate as a part of the Power Department. Its organization is shown on Fig. 1-13. The FPC Mechanical Engineering Department, composed of engineers experienced with many other power projects in the Florida Power Corporation system, as well as certain consultants available to this project are to provide all project management. Organization of this department is shown on Fig.1-1k. The FPC Production Department is responsible for the operation and maintenance of all generating plants and vill assume this responsibility for Units 3 and 4 after their being placed in commercial operation. l Gilbert Associates, Inc. has been retained by FPC as the Architect-Engineer for this project and vill assemble the necessary information for all required site studies and plot plans. In addition, they will_ furnish all plant layouts and system arangements and will draw up specifications for major equipment and systems. The firm of NUS Corporation has been retained as Environmental Consultants to assist in the preparation of reports and studies, to serve as desig? review cpecialists and to furnish guidance in nuclear related matters associated with the securing of the required permits for the project. FPC will mai.ntef_A as a part of the Mechanical Engineering Dept. its own Construction Management Staff under the Nuclear Project Manager (Chief Mechanical Engineer) to supervise and coordinate the construction of the nuclear units. This staff is experienced in power plant construction man-agement having exercised this supervision on all major plants in the FPC cystem. FPC has contracted with The Babcock & Wilcox Co. to design, manufacture, l deliver to the site, and erect the complete nuclear steam supply system. l2 In addition., B&W will supply competent technical consultation during erec-tion, initial fuel loading, testing, and initial start-up of the complete nuclear steam supply system. B&W vill also cooperate with FPC in the training and licensing of FPC operating personnel prior to and during the ) start-up and initial operating period.
1.7 CONCLUSION
S, 4 The personnel assembled to design, construct, and operate the Crystal River Plant are competent. It is their combined intention to make this a conser-vative design, and one which can be operated to produce electric power safely and economically.
$000055 1-28 (Revised 2 7-68) y e
Toward this end-- l3 a. The s!te has been examined and found to be suitable for the nuclear plant. The plant at this site is compatible with surrounding population and land uses, present and expected. Site characteristics of meteorology, 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 will be thoroughly tested to confirm that the desirable features were designed into it, and that it will perform as expected with full safety margins. ,
- c. The reactor vill be installed in an enclosure both modern and conservative in design, which will be able to contain and control all materials, vapors or energies which could conceivably be re-leased as a result of an accident under any coincident condition.
Supplementing the enclosure capability will be engineered safe-guards which will reduce to a very minimum the consequences of any accident and insure that the dynamic conditions existing after an accident are kept vell within safeguards design parameters.
- d. The plant vaste and emergency syste=s will be designed to release only effluents permitted by the AEC Regulations. Where practicable,
('/T A- liquid and gaseous vastes vill be treated so that the effluents contain a minimum of radioactivity and significantly less than that allowed by applicable regulation.
- e. A training program is planned which will adequately prepare oper-ating personnel so that they will be qualif*ed to test, start-up and operate the nuclear unit.
In consideration of the above circumstances and plans, it is concluded that the proposed Crystal River Plant Units 3 and 4 can be designed, constructed, and operated in a safe manner; that the proposed design vill provide adequate protection to the public from any sequence of events resulting in disablement of equipment from causes, natural or mechanical; and that Florida Power Cor-poration and their consultants are qualified to design, construct, start, operate, and maintain these proposed nuclear generating units in accordance with all applicable laws and regulations and in a manner satisfactoty to the Atomic Energy Co= mission, to the public interests, and to itself. t
. , S '-
00000&56 s k ): q). ' l-29 1 4
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Table 1-1 Engineered Safeguards Total Equipment Inrtalled 1 Function (per nuclear unit) High Pressure Injection 3 pumps (makeup) 1 storage tank Core Flooding System 2 tanks Low Pressure Injection 2 pumps (decay heat removal) 2 heat exchangers l2 Reactor Building Spray 2 pumps System 1 sodium thiosulfate tank Reactor Building Cooling 3 pumps Systen l1 3 emergency cooling units i 00000051 l O l 1-30 (Revised 2-7-68) 1
, r,' --
4
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Table 12 caemarteos of Destan parameters t
,,, (per statten unit beste unlese noted)
- 1Rs l Crystal aver Dree attle Islaat Ocomme shaclear Stattaa
, en ., item :
- taitt J or b Nuclear Station Unit 1 or 2 Turkey Mat R>. 3 or %
I. 1 Itriraulie and %ernal Desten Parameters y%
*~ aated neat htput, sett 2,452 2,452 2,452 2,097
.I" ' P=ted uset htput, atu/hr 8,369 x 106 8,369 x 16 8,369 x 106 7.157 x 106
, % . am overpower, 5 lb 14 in 12
. .;p e , Pressure (nominal), pain 2,200 2,200 2,200 2,250
. System Pressure (minism steady state), pela 2,150 2,150 2,150 2,220 Power Distribution Factors seat cenerated la ruel and claddinc, $ W.3 W. 3 W. 3 W.4 (nuclear) 1.85 1.85 1.85 1.75 (nuclear) ' 3 15 3 15 3 15 3 12 got 1 Factors F 3 24 3 24 3 24 3 25 Dunktio(nue. andConditions at Rated mech.) 2.27 % 3) 2.71 [W-3) 2.Ff LV-3) 1.85 (W-3) 1.60 anw-168) 1.60 Lanu-168) 1.60 i aAW-168)
. Minimum Dun ktatto at Desscn overpower 1.73 w-3) 1 73 iW-3) 173 LW-3) 1.M (W-3) 1 38 aAW-168) 138 danw-168) 138 LaAw-168) I coolant now Total F1wu hte,1b/hr 1313 x 106 1313 x 1 1313 x 1 100.6 a 1 Effective now Rate for Heat Transfer, Ib 120.9 x 106 1%9x1 120 9 x 1 91 5 x 1 Effective now Area for Best Transfer, ft b7 75 47.75 47 75 33.0 Average Velocity Along Fuel Rods, ft/see 15 70 15 70 15.70 13 9 g Avere w e ms Velocity, Ib/hr-ft2 2 53 x 106 2 53 x 106 2 53 a lof 2 35 x 106 g Coolant Temperature, F . w soninal talet 555 555 555 A6 5 , ), Maxime Inlet due to Instrumentation Error and Deedband 557 557 557 5A 5 Aversco Kise la Vessen 47.8 b7.8 67.8 %
- Average Rise in Opre 49 3 49 3 49 3 59 Average la core 379.7 579.7 579 7 51 7 Average la Yessel $78.9 578 9 578 9 574 Nominal outlet of Hot Qiannel 6M.4 6M.4 CM.4 647 2
Avere,,e Film Coefficient, atu/hr-ft -F 5,000 5,000 5,000 5,500 i Averace Film Tetprature Difference, F 31 31 31 30 8 Beet Transfer at 2001 Power Active Heat 1.ansfer Sarface Area, ft* 48,578 48,578 48.578 42 460 C Average Heat nux, stu/hr-ft2 167,620 167,620 167,620 1(1,200 g mime Issat Flux. etu/hr-fte Average %ernal OL i it, kw/ft
$3 5.4,000
%3,000 $3,000 533 / 00 54 5.4 53 C misse Thermal output, kw/ft mise clad sarface waperature at 17,5 li.5 17 5 17 3 i
3 Muminal Pressure, F 63 6% 6% 657 4 Q Puel Central Superature, F Maximum at 100', Power D 4,160 4,160 b,160 b,070 i Maximun at 111d overpower 6,600 4,400 6,400 4,270 l Q' Themal output, kw/ft at Maatam overpower 19,9 19 9 19.9 19 4 *
@ 2 Core 6'ccha31cel Des 12.1 Parameters y -
4 e e1 A.ae 11es } t D'81=" caA cea caA cea Ca,1 esa ace cantess A)d Pitch, ta.
. 0 558 0 558 0 558 0.563 1
r W 8
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Table 1-2 (Cont'd) Crystal River FAant D ree Mile Island Item Oconee hciear Station Unit 3 or 4 helear Statinn Unit I and 2 hrkey hint No. 3 or b Overall Dimenssons, in. 8.522 x 8.522 Fuel We1@ t (as UO 8.522 x 8.522 8.522 x e.522 8.426 x 8.k26 2 ), Ib 201,520 201,520 201,520 179,000 Total Weicht, Ib 283,200 283,200 283,200 226,200 Number of Grida per Assembly 8 8 8 Fuel ikxis 8 Number 36,816 36,816 36 816 32,028 Outside Diameter, in. 0.k20 0320 0.k20 Diametral Ap, in. OM6 0.b22 0.006 0.00b 0.0065 Clad 2 1ckness, in. 0.026 0.026 0.026 Clad h terial Eircaloy.% 0.0243 Zirealey.g Zircaloy b Eircaloy Fuel Pellets Material U02 sintered Density, % of theoretten1 95 UO2 sintered UQ sintered UOp sintered 93 95 94 93 Diameter, in. 0.362 0 302 0 362 0 3669 Innsth, in. 0.8 0.8 0.0 0.600 Control Rod Assemblies (CRA) Neutron Absorber 5% 01-15% In-80% Ag 5% Cd-15$ In-80% Ag Cladding hterial 5% Cd-15% In-80% Ag 5% Cd.15% In-80% Ac 304 SS . cold worked pk SS - cold worked 304 S3 cold worked Clad D ickness, 2n. 0.018 304 SS - cold worked o,olg 0.018 0.019 Weber of Assemblies 69 69 thaber of Control Rods per Assembly 16 69 41 16 16 20 Core Structure Core Burrel ID/CD, in. 147/150 167/g50 147/150 133 5/137.25 Dermal Shield ID/CD, in. 155/159 155/159 155/159 141.0/1475 3 Freliminary helear Design p t_a Structural Characteristics
^
Fuel Weight (as UOy), Ib 201,520 201,520 179,000 201,520 [ Clad Weight, Ib Core Diameter, in. (equivalent) 43,000 b),000 43,000 35,600 4 128.9 gpg,9 128.9 119 5 P Core Height, in. (active fuel) IM gg IM IM (n Reflector Rickness and Camposition G Tby(waterplussteel),in.
- 12 12 10 Bottom (water plus steel), in. 12 12 12 10 Side (water plus steel), in. 12 w 18 18 18 15 1
F8 H2 0/U (unit cell - cold) 2.97 2 97 2.97 3.48 Number of Fuel Assemblies 177 h g Fuel Rods / Fuel Assembly Performance Characteristics 208 177 208 ITT 200 157 204 v loading h chnique 3 region 3 region 3 ,,gga, 3 region Fuel Discharge Barnup, MWD />f!U Averace First Cycle 12,b60 8,260 14,000 Equilibrium Core Average 28,200 12,460 28,200 28,200 27,000 Feed Enrichments, w/o U-235 No. 3 No. 1 Region 1 2.29 Region 2 2.2 p 2.24 2.28 2.64 I3 Region 3 2 90 2.64 2.47 2.Q C Equilibrium 2 94 2 90 2 77 3 03 2 73 g Control Characteristics 2 94 .. Effective Multiplication (beginning of life) Nos. 3 and 4 C Cold, No Power, Clean 1 302 1.302 po. I 1 312 no. 2 lW 1.275 8 g ik>t, No Ptwer, clean 1.247 1.247 1.258 1.201 1.225 Hot, Rated Power, Ie and as Equ111 brims 1.158 1.158 C 1.167 1.119 1.170 -- O ' Ui D i ex O. O
- g
\ \ %
Table 1-2 (Cont'd) Crystal River Plant Three Mile Island Ocones helear Station Item Unit 3 or 4 suelear station Unit 1 or 2 Turkey Point rka. 3 or 8; control Rod Assemblias Material 5% 01-155 In-80% As 55 01-155 In-80% As 55 01-15% In.80$ Ad 55 01-15',In-905 A-IPmer of Assemblies 69 69 69 41 haber of Absoster Rada per CRA 16 15 16 20 Total Rod Worth ,$ IC .0 10.0 10.0 7.0 Boron Concentrations 3 [.e To Shut Reactor Down With' Rods Inserted
- (clean), cold / hot pp 1290/1080 1290/1080 *290/1150 2300/2300
- BoronWorth(hot), ppa 1/100 .1/100 1/100 1/130
* *
- fAkj
,. Boron Worta (cold), $1Ty,ppe 1/75 1/75 1/75 1/100
,* Kinetic Characteristice '*
kderator h mperature Coefficient, F +1.0 x 10*% to .}.0 x 10*% +1.0 x 10*% to -3.0 x 10 4 +1.0 x 10*% to -5 0 x 10*4 + 1.0 x 10*% to -3 0 x lo.h hderator Pressure o>erficf gnt, psi 1.0 x 10-6 to e5 0 x 10-6 1.0 x 10-6 to *).O x 10 6 -1.0 x 10-6 to e).O x 10*6 -1.0 x 10 6 go ,3,0 g 304 Muderator Void Coufficient, $ void +1.0 x 10-b to .3 0 x 10-3 41.0 x 10*% to .3 0 x 10*3 +1.0 x 10*% t,o -3.0 x 10*3 +0 5 x 10*3 to -2.0' x 10-3 Doppler coefficient, F 1.1 x 10*I o -1.7 x 10*5 t -1.1 x 10*3 to .1 7 x 10-5 1.1 x 10-5 o .1.7 x 10-5 t 1.0 x 10-5 to .2.0 x 10-5 p, 4 Principal Design 7krameters of the Reactor i Coolant ftretem W W % stem Rest @ tput, MWt 2,468 2,b(8 2,468 2.097 Ostem Rest oatput, Btu /hr 8,h23 x lok 8,k23 m id 8,423 x 18 7,156 x 10 6 m Ope stine Prwssure, peig 2,185 2,185 2,185 2 g215 N Reactor Inlet Temperature, F 555 555 555 5%.5 fj Reactor @ tlet h aperature, F haber of Imops
@3 603 603 6001 g 2 2 2 3 m Design Pressure, peig 2,$00 2,500 2,500 it,b85 n Design Tangerature, F 650 650 650 650
- A' Itrirostatic het Pressure (colo), pois 3,125 3,125 3,125 3,110 Coolant Volume, inclu ing pressuriser, ft3 11,800 11,800 11,800 9,800 Y 1btal Reactor Flav, sPm 352,000 352,000 352,000 W ,hoo 5 tor Coolant % stere code Requirements w
$ Reactor Vessel Steam Oenerator AfBE III, Class A A3 E III, Class A A3E III, Class A AGE III, Class A Tube Side Asg III, Class ,; ASE III, Class A A3E III, Cass A A3E III, Class A Siell Side A38 III, Class A ASE III, Class A AG E III, Class A A3R III, Class C C Pressurizer Ass III, Class A ASE III, Class A A3 E III, Class A A*HE AII, Claan A l
g , Pressuriser Beller hak Pressuriser Safety Valves A3s III, Class C ASE III, Class C A3E III, Class C A3E III, Clus C A3g III A3 s III ASE III ASE III C. Peactor Coolant Piping (mASI B31.1 lEASI B31.1 0GASI B31.1 USA 81 931.1 Cyy Reactor Coolant Pump Casing .As s III, Class A ASE III, Class A A3E III, Class A C) 6 Princ1Dal Design 7krameters of the 4)- ' Rome.tdr Vessel [C
-()
Material SA-302, Grade B, clad with Type 304 austeettic M SA-302, Orade B, clad with SA-302, Orade B, clad with Type 304 austenttia ss SA 302, Orade B, clad with Tyyn 304 austenitic S3 TYPE 304 austenitic SS 1
1 Tele 1-2 (f*ont *4 ) Crystal River Three Mile Island I Oconee Ifuelear Station iten Unit 3 or b tauclear 0tation Unit 1 or A Furkey Point Ib. 3 or b Desscn Pressure, psic 2,500 2,500 2,500 2,483 Destan Waperature, F 630 650 630 ( 30 OperatinC Fressure, psi; 2,18} P,193 2,183 2,233 Inside Diameter of Chell, in. 1 71 171 171 133.) Outside Diameter Across !Inasles, in. 2b7 Pb9 249 240/235-3/8 Overall Helcht of Vessel aryt ;1osure Ileast, f t-in. 41-8 3/8 bl-8 3/8 41-8 3/8 41 0 Ministas Clait Dickness, in. 1/8 1/3 1/8 }/32 7 I rtncipal Destu n Ihrameters of the Steam c.enerators Bkseber of Units 2 2 ? 3 Type Vertical, once-through Vertleal, once.throuch Vertical, once-throuch Vert teal, U-tube with inte-with intecral saper, with inte;ral super- with integral super- r,ral sm>1sture separator. heater. heater. henter.
' Wbe mterial Inconel inconel Inconel Inconea shell mterial Carbon steel Cebon steel carbon steel carbon steel Tube Side Dest;n Pressure, psi; 2, MO 2,500 2,500 2,485 Tube Side Dest;n Tewperature, F (,9 630 650 6A I
%be Side Dest;n Flow, Ib/hr 63.66 x 106 65.(4 x 10b 6).66 x 10 6 33.53 x 10h ,
Shell Side Design Fressure, psi. 1,op 1,0)n 1,050 1,085 Shell Side Desten Tseperature, 7 600 (G1 600 600 Operating Pressure, Tube Olde, ikveinal, ps10 2,18) 2,185 2,1% 2,2 35 Operatim Pressure, Chell Side, Extuam, pal; 9go 910 910 1,005 ,, Maximum ebisture at Outlet at Full Innd, ", 33 F superheat 35 F superheat 3) F superheat 1/4 g ltydrostatic Wst Pressure (tube side-cold), w psic 3,1p5 3,125 3,125 3,110 8 'Prinefral Desit.n Parameters of the Beactor Coolant Ihars flumber of Units g b b 3 TYP' vertteal, single stage Vertleal, sin o le stace Vertical, single stage vertical, single stage. itadial flow with bottom suction armi horizontal g Design Pressure, psic discharge. 2 %0 2,500 2,500 2,%8} C Dest;;n Testerature, F h (50 69 (-)0 g Operatinc Pressure, !;ominst, pst: 2,183 2,183 2,183 2,233 Saetton hoperature, F 555 555 555 W5 O Desten capaciti, com Destcn 'tbtal Developed f>ad, ft 88,000 88,0m 88,000 88,'800 3p 370 370 256 Rydrostatic Wst Pressure (cold), pet; 3,123 3,123 3,123 3,110 Q M tor Typ* A-C Imluction, stWe A-C Intaction, sincie A-C Imiuetion, simi, A-C Induction, single speed spee>I sp'*d
@ mtor Rating (nominal), hp speed 9,000 9,0n0 9,000 3,300 9 Prineiral Dest;n 15rra,eters of the 14eactor hlas Piping; mterial Carbon steel clad with SS Carbon steel clad with G3 (hrbon steel clad with S3 Austentile SS S 29 Hot Cold IA lec (ID),
(ID), in". in 36 36 20 3C 20 27-1/2 ! 3 V t=1 H I EQ ' m O C 15 c+ y 1 O O ~O
?"*'t _
M k ( ,/ Table 1 2 (Cont'd) Crystal River Plant Oconee helear Station Unit 3 or b Three Mile Island . . ItMs Nuclear Station Unit 1 and 2 Turkey Pcint No. 3 or % w.
.. Between Pump and Steam Generator (ID), in. 28 20 28 31 10" Reactor Ballding System Parameters
- Type Steel-lined, prestressed, steel-11 reed, prestressed.
- Steel. lined, prestressed, Steel. lined, prestrescal.
post-tensioned concrete, post-tensioned concrete, post-tensioned concrete, post-tenatoned coss. rete, vertical cylinder with vertical cyltraler with vertical cy11rster with vertical cy11aler v1'.h flat bottom ard shallow flat bottom aid shallow flat bottom ard shallow flat tottom and si.e.13cw dnsend roof. domed roof. domed roof, domed roof. Design P9.rameters Inside Diameter, ft 430 130 116 116 Ileight, ft 187 187 206 177 Free volume, ft3 2,000,000 2,000,000 1,rjo0,ooO 1,%0,000 Beterence Incident Pressure, psig 55 % $3 50 Reference Incident Energy (Eg), BLu 106,700,000 y4,700,000 306, M ,000 272,000,000 Energy Required to Produce incident Pressure (E2 ), htm 335,200,000 335,200,000 341,0u6,oua 300,000,000 Ratto; Eg/E2 0.915 0.915 u.&y-( 0.'jn? Ratto; (En - Eg)/Et 0.093 0.093 p. 13- 0.103 Concrete '161ckness, ft vertical wall Dume 31/2' 3-1/2 3-3/4 31/2 3 3 3-1A 3 Reactor Ba11dind Isak Prevention Icak-ticht penetrations Ink-ttsht penetrattore Imak-ticht penetrations leak .1;ht g eettstions and Mitigation nrd continuous steel g and continuous steel ard continuous steel anj vontlu.m t tec) g .f iner. Autamatic isola. laner. Automatic 1sula- 11:str. Automatic 1sola- line r. I.6 Wr.'.1 : f .;c l. . W tion where required. tion where required. tion where required, tica here rs gire-i. \ft A N o 4 i H - Caseous Effluent Ptarge Dischar;ed vent above top Discharge vent above tcp Discharge .ent abcve top "'t.rwct. ;.arti al:.te f1D .ra p, of Reactor 8411 ding C'f flee.ctor ita11 ding; 3r Reac'or 9411dirc erd ccattcra. Par *. cf i W (-200 ft above grade) (*200 ft above grade) (~ L O ft above ., rale) tie rain en.t.aust :/sts. l 11 Encineered Safecuards k (p fatety Injection System., Ik). of II1Ja.nlead Pumps v 5 3 3 Eb. of Iow liest Pursps 1 2 5 2 lienctor Ba1141N Duercency Coolers l, No. of Units 3 le 3 3 Air Flow Cap'y. Each, at Accident C). Condition, cfm %,000 A,WO %,000 (30,000 ' C) Core flu. Flot=11nc of Te"iksSystem 2 2 2 Total Volume, ft3 3 C3* 2,820 2.829 2.821 1.tn fa C)' Pustaccident ho. of Units Filters w, 1. kne None inside Iteactor Building. None C) Air Flow Omp'y. Euch, at Accident Cord 1 tion, cfm Isakage from penetrations C) Type collected, filtered, and dischergeo through station ("1s vent.
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hble 1-2 (Cont'd) 4 Crystal Itiver Ihree Mlle Islerut Oconee maelear Station Item Unit 3 or 4 helear Station Unit 1 or 2 1hrW Peint No. 3 or b Filtration !tedi.citon Rate nr3[.*e (tr - 0.? per pr.ss), hr* 1 Reactor Bt1111N .",Drty No. of Ptr ps 2 2 2 2 Inclu11N .ioditsi 3.focuira'a In,;c;tten Yes Yes No Ito her oency Ibwer Ceneret or ; tits, ::o. 3 Shared 3 2 2 for both Unitg hpe Diesel Diesel On-site hydroelectrie taats Diesel Eno lveral 2.fe,unrds Operable Fro'. All engineered safeguards All engineered safecuerds All eNineered safeguards 1 High head Safety 2njec- ; IAer,ency Ibuer Source (utatrasi) equipment is capable of e sulpment is capable of equipient is capable of tion (SI) pump - being operated from on. be1N operated from on- beirc operated fram on- 1 low head SI pump f site emergency power. site ener;ency power. site emergency power. 3 Containment air rectr;u-lettom unite I 1 contalissent spray pump f 1 Service water pump H I t.a Os 1 a 1 O O O O O M H E f0 m O O D ct w
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