ML20030A443

Chapter 3 to Preliminary Hazards Summary Rept for Big Rock Point, Description of Plant

ML20030A443
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	01/18/1960
From:	CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
References
NUDOCS 8101090516
Download: ML20030A443 (60)

Text

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S E C TIO N III.

DESCRIPTION OF THE PLANT A.

GENERAL FEATURES The proposed Big Rock Point Nuclear Power Plant will consist of a nuclear steam supply system, a power extraction system and as-sociated service f acilities. The principal structures are:

1.

A 130-foot diameter reactor enclosure.

2.

A turbine building housing the power extraction equipment, the control room, and other service facilities including:

2. I radioactive waste disposal facilities containing equipment for treatment and disposal of radio-active wastes;

2. 2 administration offices;

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2. 3 equipment and maintenance shops; and

2. 4 warehouse facilities.

3.

An off-shore submerged intake structure and an on-shore structure containing the circulating water pumping and screening equipment.

4.

A 240-foothigh stack for the disposal of gaseous wastes.

The boiling water reactor system proposed has forced circulation, is single cycle, and is light water cooled and moderated. Energy is removed from the reactor core by the coolant and the resulting steam-water mixture flows to the steam drum and from there the steam flows directly to the turbine control valves. Water from the steam drum is returned to the reactor vessel by the reactor recirculating pumps. The steam flows through a conventional tur-bine cycle with a condenser, feed pumps, and extraction feedwater heaters. Feedwater is returned to the steam drum.

The light water moderated and cooled reactor core is contained within a single reactor vessel. The fuel consists of slightly en-riched uranium oxide pellets encased in stainless steel tubing.

Th e individual fuel rods are grouped into fuel assemblies. This

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III-2 g

grouping of the fuel rods into assemblies provides for easier han-dling and, as the entrance to each assembly is provided with an orifice, gives a means of adjusting the pattern of flow and steam generation within the core. The assemblies are grouped vertically.

Control rods enter the bottom of the reactor core and are operated by a hydraulic system located below the reactor vessel. The con-trol rods are used to start up and chut down the reactor, control power, and to shape the neutron flux pattern.

The turbine has a maximum expected capability of approxiinately 75,000 kw, and is a tandem compound machine with special mois-ture removal features for handling saturated steam. It is con-nected to a 88,235 kva generator with hydrogen cooling. The turbine is designed for 1450 psig, saturated steam and is designed to heat the feedwater with extraction steam in three stages.

High purity of the water for the steam supply system is provided by a condensate demineralizer through which all feedwater flows, a continuous clean-up demineralizer designed as part of the nuclear steam supply system, and a make-up demineralizer to purify make-up water to this system. Maintaining high purity of the water pre-vents accumulation of any radioactive corrosion products in the system and permits normal or nearly normal maintenance procedure.

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III-3 B.

ENCLOSU RE l.

General Preliminarf cafeguard design information for the enclosure vessel and considerations of importance to an evaluation of the safety of this vessel are given below. The structure is to be a spherical vessel 130 feet in diameter. It will house the reactor and the other major components of the nuclear steam supply system.

2.

Material The material sf construction for the enclosure and all appur-tenances subject to the Code (see item 6 below) is SA-201 Grade B, Firtbox-tiuality steel produced to SA-300 specifi-cations.

L.

Design Pressure The enclosure design pressure must equal or exceed the peak i.2 sure resulting from the " maximum credible accident. "

C.. the basis of our present concept of this accident, discussed in Section IX, the desi;;n pressure will be set at 27 psig for a sphere 130 feet in diameter with a free volume of 80 per cent.

4.

Test Pressures The enclosure will be tested for structural integrity at a pressure of 1 1/4 times the design pressure. The leakage rate tests discussed in item "S",

below, will be run at a pressure approxi; 1ately equ.41 to the design pres sure, (con-ventionally the test pressure is selected so that the maximum pressure reached during the test would not be expected to ex-coed the design pressure).

5.

Leakage Rate The enclosure, together with its necessary penetrations, will be designed to provide the highest degree of leak tight-ness that it is practicable to obtain. The enclosure leakage rate will be assumed to be of the order of 0.5% per day at the design pressure, for purposes of preliminary design con-siderations.

6.

Design Codes To insure an mclosure of the highest intetrity, all provisions required for the ASME Boiler and Pressare Vessel Code ap-proval stamp are being incorporated in the design, fabrica-tion, and erection of the enclosure with special considerations as indicated below. The pertinent Code sections include the latest edition and supplements of Section II (Material Speci-fications),Section VIII (Unfired Pressure Vessels), and Sec-tion IX (Welding Qualifications).

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III-4 Significant exceptions to the Code, which have received Code approval for nuclear facilities, are as follows:

6.1 No internal pressure relief device will be installed.

6. 2 The vessel, as a whole, will not be stressed relieved be-cause of its large size; however, any plate segment, wholly containing a penetration, nozzle, or column con-nection, will be furnace stressed relieved after insertion of the penetration. If a large penetration intersects more than one shell plate and contains seams joining metal over 1. 25 inches thick other than the shell plate, that portion will be furnace stressed relieved as a unit before welding into a penetration assembly or into the shell.

6.3 No thickness allowance will be made for corrosion.

6.4 The maximum stresses from live or earthquake loads are not considered simultaneous with those produced by the maximum internal pressure.

6. 5 The standard hydrostatic test at 1. 5 times the nominal design pressure cannot be performed since the weight of the water would collapse the sphere. A pneumatic structural test at 1. 25 times the nominal design pressure will be performed, as provided for in the Code as an al-te rnative.

7.

Mis sile s No special design measures are considered necessary to protect the enclosure against any credible missile problent.

Safety of the enclosure against damage by missile effects that could conceivably accompany, a severe boiling water reactor accident has been considered in some detail. The conclusions of these considerations are, very briefly:

7.1 There is no indication intlis boiling water reactor de-sign of an energy source with sufficient magnitude to create a significant danger from missile effects. Spe ci-fic items considered in this respect were the metal-water reactions and nuclear excursions.

7. 2 In the unlikely event of a reactor rupture, the rupture would be of the ductile type not conducive to the develop-ment of penetrating missiles.

7.3 Failure of enough head-closure bolts to permit the whole reactor head to fly off and possibly break the en-closure is not considered credible.

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7. 4 If missiles were in some manner to be nevertheless

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generated, the concrete shielding and other structures surrounding the reactor would protect the enclosure shell from any substantial missiles for any plausible missile path.

8.

Direct Radiation Protection Direct radiation arising from radioactive material released to the enclosure in the event of a severe accident accounts for the majority of the radiation exposure that may be received from such a contained source. Gamma shine, that is exposure from air reflected gamma radiation, normally contributes only a very small additional exposure.

The most significant direct radiation exposure problem that may be associated with this plant would occur in connection with th e

" maximum credible accident" discussed in Section IX. Protec-tion against this problem is provided in part by the design char-acteristics of the plant which limi'. the extent of release of radio-active materials in the event of this accident. Additional protec-tion is provided by the concrete walls surrounding the reactor and steam drum. These walls effectively reduce the fraction of the released and airborne radioactive material which may be "seen" from any given point outside the enclosure. Further, there are many spaces above ground level which are enclosed by thick concrete walls which will also reduce the " gamma shine" prob-lem.

In any event, the extent of radiation exposure which may be re-ceived offsite from such a source is quite negligible.

(Reference Secdon IX-E6) 9.

Enclosure Isolation consideratio'ns The isolation provisions for the enclosure require that all nor-mally-open penetrations, including the ventilation openings, be closed complete enough and fast enough to assure that total estimated out-of-control leakage to the atmosphere.

ing a postulated reactor-rupture accident is less than 0. 05% _..

the total radioactive material that might become dispersed be-fore completion of closure. The specific features which must be considered in this respect and the means for effecting their control are presented below:

9.1 All penetrations normally open during operation are to be provided with two valves in series, if they are open to the interior of the enclosure, or to the interior of the reactor or to any portion of the primary loop, or could be broken as a result of a reactor-rupture acci-dent. At least one valve on each such line must be closed automatically in event of accident. All must be l

III-6 capable of being closed manually from the control room or from another place that would be accessible in the event of accident. (The isolation valves on the primary steam lines are not directly closed by any scram signal, but they will close automatically from low reactor pressure--a conditionthat would be encountered in any system rupture approaching the severity of the postulated

" maximum credible secident. ")

9. 2 To minimize the potential for contamination spread in the event of a major pipe rupture, the two valves in series in 9. I above, must be on opposite sides of the enclosure wall for pipe penetrations open to the interior of the reactor or to any portion of the primary loop.

9. 3 The ventilation openings are to be automatically closed by all scram signals. (This is dore to minimize radioac-tive-contaminant escape through these large sphere openings in the event of accident by assuring a " head start" for their closure.)

9.4 Emergency closure of piping penetrations are actuated automatically by only those scram signals which signify that a primary system break has occurred, or that such

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a break is potentially imminent. Signals to be utilized for such closure are high-sphere pressure and low-water level in the reactor vessel. Penetration closure will also be possible on a manual scram.

9. 5 The isolation valve in the off-gas system is designed to close automatically from an appropriate signal from the equipment monitoring the radiation level of the off-gasses.

Such a signal will not directly cause a scram, but the i

Iow vacuum resulting from closure of the isolation valve will result in loss of condenser vacuum which will initiate a scram shortly if appropriate operator action is not taken to shut the reactor down.

9. 6 The ventilation and steam line automatic isolation valves must be designed to be fail-safe, i.e., they will close upon loss of normal power or hydraulic control as the case may be.

9. 7 A single valve is sufficient for penetrations closed during reactor operation. Howeve r, such penetrations must be protected by interlocks or operating rr.es against being opened during reactor operation, or in potentially hazardous situations when the reactor is shut down.

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9. 8 The enclosure rnust be provided with an automatically-con-trolled vacuum " breaker" device to assure its allowable de-

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sign negative pressure is never exceeded.

III-7

10. Earthquake-Proof Design Features The effect of earthquake considerations on the design of the containment vessel has been investigated and it is concluded that " earthquake forces will not govern the design, since the wind force on the vessel at the design velocitj of 100 miles per hour is greater than 0. 05 times its weight" (that is greater thart the ground acceleration force of 0,05 g estimated for the site area).

11.

Shielding The principal radiations from which plant personnel must be shielded are neutron and gamma radiation leaving the reactor during operation, fission-product-decay gamma radiation from the nuclear fuel, and the gamma radiation of the nitrogen-16 created by neutron bombardment of the cooling water. Othe r sources or radiation include oxygen-19, neutron-capture activa-tion products of reactor materials, and corrosion products from the entire system.

The design of the shielding is based on the maximum permis-sible exposure rate of 5 rems per year, as well as applicable 13 week do.=e limits. The target weekly exposure rate limit is taken as 100 mrems/ week, In carrf ng out the above, the following maximum dose rates i

are established for the deangnated arcac:

11.1 Areas where access is not controlled:

0. 5 mrem /hr.

Such areas include the Control Room and adjacent areas, and outside areas a round the process buildings. Thus exposu're in such areas for a 40-hour week will not contribute more than 20% of the working limit dosage of 100 mrems and probably will average less than 10%

This will reserve most of the permissible radiation exposure of plant personnel for radiation zone entry, 11.2 In certain isolated cases where extended occupancy may occasionally be required: i. 5 mrem / hour.

11.3 Reactor Enclosure and Turbine Building areas requ;r-ing pe riodic entry for sampling, inspection, auxiliary equipment maintenance, etc. :

1. 5 to 15 mrem / hour.

11.4 Infrequently entered areas: 15 mrem / hour, or greater.

In the unlikely event ot a " max. c redible accident" which might re suit iri fission products being dispersed within the enclosure, the con-trol room will be sufficiently shielded to reduce the exposure to less than 0. 5 rem during the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. This will enable

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personnel to occupy the control room to take emergency action to limit the magnitude of the accident. Provision will be made so that, 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after such an accident, personnel can leave the plant.. wit hout receiving more than 25 rems, the maximum per-missible emergency dose.

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C.

PLANT LAYOUT The figures on the following pages show the site and tentative layout of the plant.

l Figure III-l Site Plan Figure III-2 Reactor Enclosure Sphere Figure III-3 Operating Floor Equipment Location 1

Figure III-4 Ground Floor Equipment Location Figures III-5, III-6 Enclosure and Turbine Building Sections k

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III-15 i

D.

NUCLEAR STEAM SUPPLY SYSTEM DESIGN 1.

General The design of the nuclear steam supply system will take advan-tage of the design and operating experience of the Experimental Boiling Water Reactor at the Argonne National Laboratory and the Vallecitos Boiling Water Reactor at the Vallecitos Atomic In addition, the results of much of the research, l

Labo rator y.

development and testing programs which were conducted for the design of the Dresden Nuclear Power Station are applicable.

The nuclear steam supply system is of the wingle-cycle, forced circulation type employing a light water moderated and cooled boiling water core.

The coolant enters the bottom of the reactor vessel and passes upward through the core. The water enters the bottom of each fuel assembly, passes through the fuel assembly orifice and the space around the individual fuel rods and inside the ente r s channel around the fuel bundles. The water passes up, along I

the fuel rods where bulk boiling produces steam. The steam-water mixture then enters the space above the core. From there, the steam-water mixture Icaves the vessel through the outlet nozzles and passes to the steam drum. In the drum the steam is separated from the water and is directed to the turbine con-trol valves. The hot water removed in the steam drum is re-turned to the reactor vessel by means of the reactor recircu-lating pumps.

The high pressure steam and feedwater piping are shown in Figure III-7 with the legend of symbols shown in Figure III-8 on the following pages.

2.

Reactor Core

2. I Core Geometry and Structure The reactor core (see data sheet on page 111-28, and Figure III-9) occupies approximately the volume of a right circular cylinder 75-3/4 inches high and 61. 8 inches in diameter. It contains 120 square fuel as-semblies and 32 control elements in a square array.

The fuel assemblies are supported frora below by the core support plate and core support grid and guided f

at the upper enda by the upper grid guide.

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CONSUMERS P'0WER CO. BIG ROCK POINT NUCLEAR POWER PLANT FI G UR E 111 -9 P00R ORIGINAL

III-19

2. 2 Fuel Assemblies The fuel assemblies are made up of 64 fuel rods in square array (8x8), positioned by spacers in the center and the plates at the ends, and having an active fuel length of 72 inches. Provision is made in design for variations in rod length arising either from manufacturing variations or-relative thermal expansion. Each fuel assembly is sur-rounded by a fuel channel which adds to the stiffness of the assembly, guides the control elemente and separates the coolant in the fuel channels from that flowing around the control elements. This permits individual orificing of each fuel a ssembly.

The fuel rods consist of stainless-steel tubing (O. 372 inch 0. D. and 0,024 inch wall) containing fuel pellets of sintered UO, O. 318 inch in diameter. The tubing 2

ends are plugged with stainless-steel fittings which lo-cate the rod in the spacers or tie plates.

2,3 Control Rods and Their Drives The 32 control element poison sections are in a square array on a 10-inch pitch and are of cruciform cross sec tion. The poison section is made of boron stainless steel with a blade width of 7-7/8 inches and length of 73-3/4 inches. The 16 interior control elements have

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cruciform zircaloy followers above the poison section to reduce the thickness of the inter-assembly water gap and the neutron flux peaking in that region with the control elements withdrawn.

j 2.3.1 Control Rod Drive Mechanism

~

The system is designed using a fail-safe philosophy such that loss of electrical power on the solenoid valves, and loss of air pressure on the scram valves initiate and/c: t<.nd to i: itiate reactor shutdown. An individual control rod drive mechanism is used to position each of the control rods. The drive mechanism is hydraulically driven for both normal and scram operation. All rods are scrammed for emergency shutdown. The control rods enter the bottom of the core to achieve efficient fuel utilization and power distribution in the boiling water reactor. Inserting the rod upward into the reactor core decreases reactivity; withdrawal or lowering a rod increases reactivity.

I

III-20

(

2.3,2 Hydraulic System Description Normal operation of a drive is controlled by external hydraulic circuits utilizing filtered unheated demineralized reactor feedwater maintained at a constant differential above reactor vessel pressure as the " drive water. "

The " insert" positions of the pressure switch-ing valves for a single drive directs drive water below the piston while the water abeve the piston is discharged to the reactor vessel. The pres-sure difference across the piston inserts the control rod into the core. Rod motion is ar-rested by returning the pressure switching valves to the "off" position. The rod will then move down by gravity until the drive locking mechanism engages a locking groove in the piston shaft and prevents further withdrawal.

The " withdraw" position of the pressure switch-ing valves directs drive water above the piston and to the locking mechanism, thereby releasing it, while the fluid below the piston is discharged to the reactor vessel, thus withdrawing the rod.

Returning the pressure switching valves to the "off" position isolates the drive from the hydraulic system and allows the locking mechanism to engage the piston shaft, preventing further rod withdrawal.

Normal rod insertion and withdrawal speeds are controlled by drive water pressure and by orifices in the hydraulic circ ~uit.

The relaxed or de-energized condition of the pressure switching valves is "off" or " insert" depending on the valve function.

A small amount of water at a low pressure dif-ferential above reactor pressure (insufficient to move the drive) is continually bled through the drive for cooling purposes.

Scram or emergency insertion of rods is accom-plished by opening scram valves which direct

" scram water" at or above reactor pressure be-low the piston while the fluid above the piston is discharged to a reservoir initially at atmospheric pressure. The unrestricted flow of water at a large pressure differential rapidly inserts the

[

rod into the core. As in normal insertion, un-intentional withdrawal of the rod after scramming m.

.-,a

III-21 is prevented by the locking mechanism. Sc ram

(

water is supplied by either the reactor vessel or by an accumulator which is kept charged with reactor feedwater. A shuttle valve in the scram circuit automatically selects the higher of the reactor vessel or accumulater pressures and directs it below the piston. Two non-adjacent drives are supplied by a single accumulator.

The scram reservoir is a closed vessel of suf-ficient volume so as not to develop a large back pressure during scram, while limiting the water discharged from the vessel after scram. The de-energized or relaxed position of all motor operated valves is in the scram condition. The system is also monitored to prevent control rod removal on low accumulator pressure.

2. 4 Heat Transfer and Fluid Flow Considerations Two principal design limits are (1) detrimental effects of high temperatures on material properties, and (2) thermal burn-out of the fuel rods.

On stainless-steel jacketed fuel rods, with diameters in the range of 3/8 inch, the limit on heat flux from the ma-terial temperature standpoint is imposed by the center fuel melting temperature. Since UO melts at around 5000* F 2

and since the rate of fission gas release from the fuel ma-trix to the gap area between the clad and fuel may be signi-ficantly higher with molten fuel, present design practice limits the surface heat flux on the hottest rod at the peak of overpower condition to a value which would not cause i

fuel melting. Since this restriction is significant only at the hottest point in the reactor, the reactor power level is limited also by the uniformity of power distribution.

Thermal burn-out can occur when so much steam is generated at some point on a heat transfer surface that it completely blankets the surface. In general, the greater the local steam fraction the lower the value of heat flux which would cause burn-out. Cons e quently, the hotter fuel assemblies, those in the core center, re-quire more fluid flow than the peripheral assemblies where the power generation rate is lower. In order to accomplish this, the outer fuel assemblies contain flow restricting orifices of various diameters arranged so that the flow pattern tends to follow the power generation patte rn.

The core and fuel design is such that at the

[

III-22

(

maximum overpower condition the minimum local ratio of burn-out heat flux to local heat flux is 1. 5.

This point of closest approach is usually located between the point of maximum local heat flux and the channel exit (where the burn-out flux is lowest).

With the optimum orificing pattern and power distribution determined, the coolant flow through the fuel channels is not permitted to go below that which will meet the above criteria.

Sufficient coolant flow is permitted in the control element spaces to prevent steam generation between fuel channels.

3.

Steam Supply System 3.1 General The major components of the steam supply system are the pressure vessel, steam drum, recirculating pumps, and auxiliary equipment. These components will be de-signed for system working pressures up to 1500 psia.

Auxiliary equipment in the steam supply system will include reactor liquid poison equipment, reactor clean-up system equipment, reactor shutdown cooling equip-ment, and reactor emergency cooling equipment.

3. 2 Pressure Vessel The reactor pressure vessel, diameter will be sized to accommodate a core which, based on present technology, will be capable of supplying sufficient steam at 1050 psia to generate 50,000 % poss. It is expected that future technology will permit this vessel size to accommodate a core which will be capable of delivering 75,000 kwe gross at 1500 psia.

The reactor vessel will be approximately 9 feet inside diameter by 32 feet in length. The base material will be high strength alloy material such as ASTM A302 Grade B firebox or equivalent. All surfaces exposed to coolant will be stainless steel clad. The vessel will be designed, built and tested in accordance with the ASME Boiler and Pressure Vessel Code. The operating pres-sure of the vessel will be 1500 psia and the design pres-i l

sure will be determined from analyses of pressure l

l fluctuation arising under various transient operating

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conditions.

~~

III-23 The vessel will contain and support the reactor coolant, g

fuel and internal reactor components.

The internals of the reactor include:

the fuel assembly supports, the thermal shield, the diffuser basket, the poison sparger and the core spray sparger.

The fuel assembly supports guide the individual fuel as-semblies which make up the core. The thermal shield prevents exposure of the vessel walls to excessive fast neutron flux levels. The diffuser basket properly dis-tributes the water entering the bottom of the core through tb.. inlet pipes from the reactor recirculating pumps. The poison sparger is used to insert poison into the core, if such a requirement should arise. The core spray sparger is used to spray the core with water in the event the core coolant ehould be lost.

In addition to housing the core, the vessel also supports the control rod drives and the control rods. The drives will be supported at the bottom of the vessel and insert the rods into the core from the bottom. Nozzles will be provided at the bottom of the vessel to house the drives.

t Water will enter the vessel from two inlet nozzles located at the bottom of the vessel. A mixture of water and steam will leave the vessel through six nozzles located near the top of the vessel. In addition to these nozzles, there will be nozzles for instrumentation, the poison system, the core spray system, and the unloading system.

Access into the vessel for refueling, changing control rods and servicing of instrumentation will be performed through a removable head closure or ports at the top of the vessel.

3. 3 Main Steam Drum A single steam drum is provided to separate the steam generated in the reactor from the steam-water mixture.

It also provides adequate storage capacity to accommodate pressure surges between the reactor vessel and the drum during transient conditions and also assures a positive sup-ply to the recirculating pumps and is located high enough to j

maintain natural circulation in case the recirculating pumps are inoperative.

III-24

(

The drum will be a carbon steel or low alloy steel pres-sure vessel with stainless steel cladding on all internal surface s.

The drum will be provided with moisture separators and dryers and will be designed and con-structed in accordance with the ASME boiler codes.

The steam-water mixture enters the drum through six risers. The steam leaves the drum through four off-takes from the drum which are joined to one 12-inch steam line which goes to the turbine. The water in the drum is returned to the suction of the recirculating pumps through four downcomers, two per pump.

3. 4 Reactor Recirculating Pumps Water is returned to the reactor vessel from the steam drum in two loops by two vertically mounted recircu-lating pumps. The pump motors will be designed for operation at 2300 V., and will provide sufficient horse-power to circulate a flow of approximately 16,000 gpm per pump.

3. 5 Main Steam Relief and Bypass Valve The turbine steam bypass valve will open automatically on build-up of reactor steam pressure above regulator setting. The bypass valve is also arranged for remote manual operation from the control console. As the steam bypass valve is opened, desuperheating water spray valves inject condensate into the bypass line to the condenser. Provision is made for accommodating bypass steam in the condenser at reduced pressure, however, no additional condensink surface is provided for this function.

3. 6 Reactor Liquid Poison System The liquid poison system serves as an emergency backup to shut down the reactor and hold it subcritical in the event that the control rods are unable to do so.

The poison system, consisting of a storage tank contain-ir.g sodium pentaborate solution, is normally pressurized to 2000 prig by bottled nitrogen. The storage tank tem-perature is maintained by electrical heaters. The system i

is initiated manually by remotely operating a valve in the

(

supply line from the tank to the reactor vessel, permitting the solution to flow into and mix with the reactor coolant.

--nm--

III-25 f

3. 7 Reactor Cleanup System It is necessary to remove the corrosion products orig-inating in the feedwater heaters, reactor vessel, recir-culating loop, piping, and shutdown system equipment to maintain reactor water purity. The reactor cleanup sys-tem sufficiently removes impurities to maintain the dissolved solids content of the reactor water below 0. 5 ppm.

Water from the cleanup pump discharge flows through the tubes of the regenerative heat exchanger where it is cooled to approximately 250 *F by et unter-current exchange with water being returned to tie reactor from the cleanup demineralizer. The stream is. pumped through a non-regenerative heat exchanger for cooling to 110 *F by exchange with reactor cooling water, and then through the cleanup demineralizer. Water leaving the demineralizer passes through the shell side of the regenerative heat exchanger and re-enters the reactor circulating pump suctions at about 485 *F.

A remote manual flow control valve is supplied on the cleanup pump discharge to permit adjustment of the cleanup i

system flow rate.

Spent resins from the demineralizer are not regenerated, but are discharged to the radioactive waste system resin disposal tank for storage prior to ultimate disposal. A fresh resin addition hopper is provided for sluicing of fresh resin charges into the system.

The reactor cleanup system is shown in Figure III-10.

3. 8 Reactor Shutdown Cooling System Removal of reactor decay heat and reduction of reactor pressure immediately following shutdown of the turbine and the reactor ic accomplished normally by controlled reactor steam flow to the main condenser through the turbine bypass line.

After initial cooling of the reactor system, the reactor shutdown cooling system is placed in operation, further reducing the reactor coolant temperature to 125 'F.

This system is in continuous operation during refueling oper-(

ations and at such other times as the reactor is shut down with irradiated fuel in the core and the head removed or loose.

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III-27 The reactor shutdown cooling system is shown in Figure III-10.

3. 9 Reactor Emergency Cooling System An emergency cooling system is provided to dissipate reactor decay heat following scram or shutdown in the event that the normal heat sinks are unavailable. This system is designed to maintain the reactor in a safe condition when all plant auxiliary power has been lost.

The arrangement of the system is such that steam generated in the reactor is condensed in an emergency condenser and the condensate returned by gravity to the reactor.

The emergency condenser consists of condensing tubes passing through an atmospheric tank of water. The tank capacity above the tubes is sufficient for removal of the de-cay heat generated in approximately 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following scram.

1 Operion of the system is initiated automatically or man-ually by opening the drain line valves, allowing condensate l

to drain from the tubes back to the reactor. Steam is then condensed in the tubes, causing heating and evaporation of the water in the tank, and the condensate continues to flow l

by gravity back to the reactor. The steam generated in the tank is non-radioactive and is discharged to the atmos-phere.

A water line to the tank permits makeup of cooling water if it becomes necessary to extend operation of the emer-gency condenser beyond 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Makeup water for this condenser is normally supplied from the condensate stor-l age tank by the motor-driven makeup pump. In the event i

of continual outage of plant auxiliary power makeup can i

be supplied by an engine driven emergency makeup pump.

3.10 Reactor Core Spray System An emergency core spray system is provided to prevent significant fuel melt down following scram and loss of all other coolant measures. The system will introduce cool-ing water through a sparger ring located above the reactor core. The system will be designed for automatic initiation from either low reactor pressure or low water level in the reactor.

(

III-28 4.

Reactor Lesign Data Preliminary characteristics of reactor design include the fol-lowing quantities in reference to rated conditions of 50 MWe gross. These quantities are not to be considered final, but are chosen as an example of,what the final design could be.

TABLE III-l PRELIMINARY DESIGN DATA (a) Reference Rating Gross Electric Power (Approx. )- MW 50 Reactor Power (Approx. ) - MW.

156 Reactor Pressure, - psia 1050 Feedwater Temp.

'F 344.7 (b) Fuel & Core Assembly Fuel Material UO2 Initial Enrichment - % U235

3. 9 Weight of UO2 in Core - lb s.

16,500 Fuel Pellet Diameter - in.

.318 Clad Thickness (Stainless Steel) - in.

.024 Fuel Rod Active Length - in.

72 Number Fuel Rods Per Assembly 64 Number Fuel Assemblies In Core 120 Circumscribed Core Diameter - in.

67. 1 Equiv. Core Diameter - in.

61.8 Over-all Length of Fuel Zone - in.

75-3/4 (c) Control Elements k Followers Number of Control Rode 32 Material, poison section 2%B-98% S. S.

Pitch (Square Array) - in.

10 Active Length - in.

73-3/4 Shape Cruciform Width - in.

7-7/8 Thickne s s - in.

3/8 Followers - number 16 Blade Width - in.

4 Thicknes s - in.

3/8 Length 56-3/4 Material, followers Zr r

=

III-29 t

(d) Heat Transfer & Fluid Flow 6

Rated Heat Generation - Btu / hr 532.4x10 2

3 Average Heat Flux at Rating - Btu / hr ft 118.5x10 Maximum Heat Flux at 125% Rating -

3

_ Btu /lir-ft2 444 x 10 Total Heat Transfer Area - ft2 4490 Fuel Center Temperature at Rated Power

• F 3840 Maximum Fuel Center Temperature at Overpower "F

4655 Peaking Factors Axial 1.43 Radial

1. 6 Local

1. 25 Manufacturing Variations 1.05 Over Power Trip

1. 25 Maximum Peaking Factor at Over Power 3.75 j

Power Density - kw/1 45 Power Per Average Fuel Rod ( Approx. ) - kw 20.3 Power Per Central Fuel Rod ( Approx. ) - kw 39 Steam Flow Rate - lbs /hr-609,000 Steam Quality Core Exit

5. 2%

Steam Quality Eutering Drum 5.96%

Pressure at Turbine - psig 1000 erature

• F 552.3 Reactor Operating Temp / channel Coolant Flow Area - in 11.79 Coolant Velocity, Average Channel - f t/ sec.

6.03 6

Recirculating Flow Rate

.lbs/hr ll.78x10 Hydraulic '>iameter - in.

.819 (e) Reactor Vessel inside Diameter - in.

106 Inside Length - it.

32

[

III-30

(

E.

POWER EXTRACTION SYSTEM -

1.

Turbine Generator System The. turbine is a 3600 rpm, tandem-compound, double flow, condensing unit direct connected to a hydrogen cooled genera-tor which in turn is connected through a reduction gear to an air-cooled exciter. Three points of extraction for feedwater heating are provided.

The turbine is rated at 54,500 kw at 1000 psig, O degrees final superheat and 3-1/2 inches of mercury absolute exhaust with 3 percent makeup allowance and three feedwater heaters in service. Also the turbine-generator is capable of operating continuously at 1450 psig, O degrees final superheat and 1-1/2 inches back pressure with a maximum expected output of 75,000 kw.

The 13,800 volt, wye-connected generator is rated 70,588 kva, O. 85 power factor, 0,80 short circuit ratio at 30 psig hydro-gen pressure. If the expected maximum output of the nuclear steam supply system is achieved, the generator rating will be increased to 88,235 kva, 0. 85 power factor, at 0. 64 short circuit ratio.

Besides conventional design criteria, all modifications nec-essary due to use of steam from a boiling water reactor are incorporated. Particular attention is given in the design of the machine to the elimination of pockets or crevices in which radioactive material might lodge. Each turbine stage is drained, either internally or externally. The turbine is provided with moisture removal buckets ahead of each ex-tration point; in addition, two centrifugal type moisture sep-arators are provided in the cross-over between the high-pressure and low-pressure sections. Materials used in the construction of the turbine are selected to minimize the wear caused by wet, oxygenated steam.

The flow paths through the turbine are shown in Figure 111-11 and extmetion drains and vents in Figure III-12.

2.

Condensing System The main condenser is designed to perform the following functions:

2.1 Condense steam exhausted from the turbine to obtain the desired vacuum,

2. 2 De-aerate the condensate, heater drains, and other re-tu rn s.

(

2. 3 Serve as a heat sink for excess reactor steam dumped

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. through the turbine bypass valve.

2. 4 Detain condensate in the hotwell to permit decay of short-lived radioactivity.

2 The main condenser is a fabricated steel, horizontal, single pass, divided water box de-aerating type unit 2

of conventional construction. It is solid supported, 7

with a rubber expansion joint provided between the con'-

denser neck and the turbine exhaust flange. The unit has an effective surface of 25,000 square feet. The condenser is located directly beneath the low pressure turbine with tubes' perpendicular to the turbine center-line.

Provision is made for accommodating bypass steam in the condenser at reduced pressure. (No ad-ditional condensing surface is provided for this func-tion,' however ) The bypass steam is desuperheated with water from the condensate pump discharge.

-In order to minimize the amount of copper corrosion products and consequent radioactivity in the system, aluminum condenser tubes are used.

A 4000 gallon, oversized, storage-type hotwell is provided to allow decay of short-lived radioactivity.

The notwell is divided by a partition plate parallel to the tubes to facilitate location of tube leaks.

A single twin-element., two-stage steam-jet air ejector with surfac e type inte r and after --condenser is provided.

Each element is capable of remov:ng 10 cfm of free dry air leakage plus 1.1 lbs/hr of hydrogen and 8. 3 lbs/hr of oxygen gas from the reactor.

A motor-driven, wet-type, rota ry vacuum pump with a capacity of approximately 600_ cfm of air at 15" Hg absolute pressure is provided for rapid evacuation of the condenser steam space at startup.

The air and gas removal equipment discharges to the main exhaust stac k through oversize piping which pro-vides holdup en route for a m2nimum of 3 minutes.

l 3.

Condensate and Feedwater System Two half-capacity, vertical, multi-stage contrifugal pumps

. pump the condensate from the hotwell through the condensate system' to the suction of the reactor feed pumps.

The condensate pumps deliver condensate through the air ejector inter and after condenser, turbine gland seal con -

I denser, condensate demineralizers, low pressure feedwater s.,

III-34 heater, and intermediate pressure feedwater heater, in series (See Figure III-13).

The low pressure and intermediate pressure heaters are of the horizcntal-mounted U-tube type with removable bundles, inte gral drain coolers, and bolted head covers.

Incoming turbine drains _ to each heater flow first into a small tank for separation. The steam then flows to the condensing zone of the heater while separated water flows directly to the drain cooling section, thus minimizing the impingement of water on the heater tubes.

The feedwater system consists of that part of the circuit between the feed pump suction and the feedwater inlet at

- the reactor, and includes the high pressure feedwater heater (See Figure III-7). The high pressure heater is of the hori-zontal U-tube type with integral drain cooler. Channel connec-tions are welded and tube maintenance is performed by cutting and removing a skirt section on the shell.

Two half-capacity feed pumps, taking suction directly from the condensate system, discharge feedvater through the high pressure heater ac.d through a common header to the reactor circulation loop. They are horizontal, multi-stage, centri-fugal pumps. The shaft seal arrangement precludes any sig-I nificant external leakage.

4.

Condensate Demineralizer System Three half-capacity mixed-bed ion exchangers designed for a maximum flow rate of 50 gpm/sq. ft. are provided for re-moval of reactor solids carry-o,ver and turbine-condenser system corrosion products from the full condensate flow.

In normal operation two of the demineralizers will clean up the full condensate flow, while the third is being rg;merated or on standby.

An external resin regeneration system consists of cation and anion resin regeneration tanks and a regenerated resin storage tank. Spent resins are hydraulically sluiced from the demin-eralizers, to this system, where they are separated, individu-ally regenerated and stored.

Spent resins are monitored for radiation level after removal from the ion exchangers before classification and regmeration.

If resin is found to contain a concentration of radioactive mat-erial so great as to poee unwarranted handling and disposal problems with the regeneration waste solution the resin is sluiced directly to the radwaste system resin disposal tank l

and replaced with a fresh resin charge. (Refer to Waste Dis-(

posal Control Systens page 111-53). Resin of low activity level is regenerated

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5.

Service Cooling Water Systems Two separate cooling water systems are provided for Big Rock plant cooling services. These are designated service water system, and reactor cooling water system.

5.1 Service Water System The service water system is an open system in which -

strained water is supplied from pumps in the intake structure and returned to the lake along with the dis-charge from the circulating water system. The ser-vice water system removes heat from the following equipment:

Generator Hydrogen Coolers Turbine Lube Oil Coolers Feed Pump Bearings and Oil Coolers Air Compressor Aftercoolers and Jackets Miscellaneous Sample Coolers Air Conditioning System Reactor Enclosure Space Heating and Cooling Two full-capacity, vertical, submerged type service water pumps are provided.

5. 2 Reactor Cooling Water System The designated reactor cooling water system will be a closed loop with water from the reactor cooling water return tank pumped through heat exchangers to supply the following equipment:

Reactor Shield Cooling Coils Reactor Cleanup Non-Regenerative Heat Exchanger Reactor Shutdown Heat Exchanger Fuel Pit Cooling Water Heat Exchanger j

Miscellaneous Sample Coolers The return header will be monitored to indicate and alarm excess radioactivity buildup. Two full capacity, horizontal motor driven pumps will be provided for the system.

(

III-37 6.

Circulating Water System

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Condenser cooling water is drawn from Lake Michigan through submerged pipe extending out from the shoreline. The sub-merged line empties into the intake structure near the shore, where the water passes through the bar racks and screens and is then pumped through underground pipes to the conden-ser by two half-capacity, vertical, mixed flow, wet-pit type pump s.

The circulating water is carried from the condenser through an underground pipe to the discharge headworks at the shoreline.

The intake structure consists of two compartments, each with a sloping bar rack, a traveling water screen, a screen -

wash pump with basket strainer, service water pump and a circulating water pump. The fire system jockey pump and emergency diesel fire pump are also located at the intake structu re.

Stop logs provide for dewatering of the screen and pump wells. An enclosure, which can be removed for servicing, is provided over the pumps and screen wells.

7.

Instrument and Service Air Systems Two full-capacity motor-driven air compressors supply both instrument and service air to the plant. They deliver air at 100 psig to a common header which discharges to the instrument air and service air receivers. Air from the instrument air receiver passes through moisture removal equipment to the instrument air supply header.

8.

Makeup Water System Makeup water to the steam and condensate system, and de-mineralized water for the reactor cooling water system and other requirements are supplied by a single mixed-bed ion exchanger of standard commercial design.

Operation of the demineralizer is manually initiated as de-termined by demineralized water requirements.

Two horizontal ce: rifugal pumps, one motor driven and one engine driven, a re provided toaipply demineralized water from the condensate storage tank to the reactor cleanup and waste demineralizers for sluicing of spent ion exchange resins to the disposal facilities. The pump also supplies rinse and dilution water to the makeup demineral-izer, makeup wate - br the emergency condenser, fuel pit, quench tank and, in addition, provides wash water for equipment decontamination by means of hose and wash connections in selected areas.

During makeup demineralizer maintenance outages, water for makeup may be obtained either from recovery of demin-eralized aqueous wastes in the waste treatment facilities,

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or by bleeding lake water into the condenser via the waste treatment system demineralizer.

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1 l

III-38 F.

MATERIALS OF CONSTRUCTION g

In general, stainless steel is used on all surfaces in contact with reactor water during plant operation. The reactor vessel, steam drum and pumps will be stainless steel clad, while the pipes in the nuclear steam supply system will be solid stainless steel.

The other components, turbine-generator, condensate system, feedwater system, etc., will use conventional materials for the rated steam temperature and pressure.

Studies will be made to evaluate the use of other materials in any area (such as zircaloy channels), where special considerations should be made.

e 1

III-39 G.

CONTROL AND INSTRUMENTATION i

1.

General The main control room will contain all the control and instrumentation essential to the operator during normal and abnormal plant operation. The operators console contains. all controls, indicators, and recorders needed-by the operator on a continuing basis for startup, shut-

= down, and operation of the plant. T'e recorder board and switchgear panels adjacent to tL-console contain controls, indicators, and protective devices for auxiliary -

plant equipment such as the auxiliary electrical system, and power plant auxiliaries. The recorder board and switchgear panels contain instrumentation which must be referred to on a non-routine basis but should be reason-ably accessible to the operator. Local control panels are provided in some cases.

Instrumentation and controls for the turbine-generator, feedwater heaters, and other portions of the power plant which are like fossil-fueled power plants are conventional.

They include normal safety features such as turbine over-speed trip, protective relaying and control, and indication and v ording of critical temperatures, pressures, and flow -

it The p.

Strument board and the control console are located 1 control room near the turbine operating

.m floo r.

Al_ supervisory and recording instruments are located on the instrument board, which also houses the amplifiers and power supply units for the reactor neutron monitoring system. The rear face of the instrument board is used to mount the process and area radiation monitoring recorders, together with other equipment, such as the generator relays, which do not require front board supervision.

The following paragraphs describe the control and instru-mentation systems for the reactor and for personnel or plant protection.

i 2.

Powe r Control The control rods are manually set by the operator, and the setting determines the reactor power level, void volume and void dis tribution. The operator adjusts the setting as neces-sary to compensate for fuel burnup and changes in xenon poisoning.

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i III-40

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The turbine control mechanism includes a conventional governor and a reactor pressure regulator. During normal operation, the turbine admission valves are controlled by the reactor pressure regulator; and the turbine speed gov-ernor output signal is some amount above the pressure reg-ulator and, therefore, not in control of the turbine admission valv e s.

The steam bypass valve is normally closed, and all reactor steam flow is through the turbine.

- With the control system set up in this manner, the turbine follows the reactor output rather than system load changes.

4 However, the necessary protection against turbine overspeed is still maintained.

3.

Turbine Protection Devices If the turbine should overspeed due to sudden loss of elect:ical load, the speed governor signal will override the reactor pres-sure regulator signal which normally controls the turbine ad-mission valves. The admission valves close sufficiently to maintain satisfactory turbine speed. This causes the reactor pressure and reactivity to rise. The turbine bypass valve is opened by means of a pressure signal and dumps steam to the condenser. In order to protect the condenser when steam is dumped from the bypass valves, a desuperheating water spray is initiated. A turbine emergency overspeed governor is pro-l vided as a backup for the speed governor.

4.

Control at Startup Upon control rod withdrawals, reactivity is increased from the cold shutdown condition until the reactor is brought to critical.

After the initial critical, the rods are withdrawn until a period of about 30 seconds or longer is obtained. This period is main-tained until the proper heating rate is established to bring up the temperature of the reactor at a prescribed rate.

5.

Reactor Protection System The function of the reactor protection system is to protect equip-ment, plant, and personnel by scramming the control rods, and in some cases, close the enclosure penetrations in the event of an unsafe or potentially unsafe condition. A functional outline of this system is shown in Table III-2.

(

l III-41 i

8 TABLE III-2 REACTOR SAFETY SYSTEM FUNCTIONS Start Close Em e~r-Close Sc ram Vent gency-Enclosure Sensors Reactor Ducts Cooling Penetrations High Neutron Flux X

X Short Reactor Period X

X High Reactor Pressure X

X X

Low Water Level in Reactor Vessel X

X X

l Low Water Level in Steam Drum X

X

'C1'osure of Steam Line Backup Isolation Valves X

X X

Closure of Turbine Stop and Bypass Valves X

X Closure of Recirculation Water Line Valves X

X High Condenser Pressure X

X High Sphere Pressure X

X X

(

Loss of Station Power X

X X

X X

X Manual Acticn X

X P90R ORIGINAL

III-42 I

The system consists of two independent, " fail-safe" safety channels which must both he de-energized to produce a scram or other safety system function. The failure of a single com-ponent or power supply does not prevent a desired scram or cause an unwanted scram.

Certain sensing elements are continuously monitored so that an operation or failure is clearly indicated and identified for quick and easy maintenan. a.

Wherever practical, the two channels are physically separated and clearly identified so as to minimize the possibility of main-tenance personnel causing an accidental shutdown.

5.1 Scram Conditions The Unit conditions which are monitored by the reactor protection system and used to cause a scram are:

5.1.1 High Pressure in Enclosure A differential pressure greater than abou* ;

{

psig between the inside and the outside of the reactor enclosure could indicate a major rupture within the enclosure.

5.1. 2 Low Water Level in Reactor This could in'dicate a loss of water so great that adequate circulation would be jeopardized or the reactor core would be uncovered.

5.1. 3 Low Water Level in Drum This would indicate a loss of feedwater and eventual low water level in reactor.

5.1. 4 High Pressure in Reactor This could indicate trouble in the reactor system. It is set so that normal pressure transients do not cause a shutdo.vn.

5.1. 5 High Neutron Flux

(

This would indicate a reactor output in excess of the safe level for continuous operation.

III-43 5.1. 6 Short Period l

This would indicate an excessive rate of rise of reactor power.

5.1. 7 Closure of Steam Line Backup Isolation Valve Closure of this valve initiates both reactor scram and the emergency condenser. The emergency condenser is brought on imme-diately to permi+. a head start on pressure reduction as a high system pressere would certainly follow such closure during operation at rated condit2ons.

5.~1.8 Simultaneous Closure of Recirculation Water -

line Valves.

[

Closure of these valves would prevent coolant circulation to the core and a prompt scram is required to prevent development of excessive fuel temperatures.

5.1. 9 Closure of Turbine Stop and Bypass Valves Simultaneous closure of these valves would result in a rapid increase in system pressure if the reactor were operating at rated condi-tions. However, in this case, either of the valves may be opened quite quickly by manual action thus limiting pressure buildup. Thus, the emergency action is limited to scram of the reactor.

5.1.10 High Conderiser Pressure The reactor w!!! he scrammed at a nominal increase in pressure and at a higher setting the bypass valves will be closed. These precautions are taken to prevent rupture of the turbine load exhause diaphragms.

5. 1. 11 Loss of Auxiliary Power The safety system is supplied from the a-c auxiliary power system through isolating motor-generator sets which have enough energy stored in their flywheels to carry them through power system disturbances.

If power is unavailable, the reactor will be scrammed and put into a " safe" condi-tion.

(

III-44 ff 5.1.12 Manual Trips Manual trips are provided for the devices which are actuated from the protection system so that the operator is able to take rapid action to protect the reactor in case of some unusual or unforeseen emergency.

5. 2 Penetration Closure Conditions The following conditions close all significant normally open enclosure penetrMions:

5.2.1 High pressure in enclosure.

5.2.2 Low water level in reactor.

5.2.3 Manual penetration-closure trip.

5.2.4 Loss of auxiliary power.

5. 3 Emergency Cooling 1,

Emergency cooling is actuated manually and from the high-pressure in-reactor sensors which also scram the reactor.

5. 4 Rod Withdrawal Limit Conditions In order to insure that tile protection system controls are set properly during the star up period, withdrawal of control rods is prevented when certain settings are imp rope r.

The following conditions prevent control rod withdrawal:

5.4.1 Setting the flux level trips to more than

1. 5 decades of the operating point during startup or refueling provides protection against an excursion caused by a localized region within the reactor. Such an excur-sien could have a flux pattern which differed appreciably from the more flattened distri-bution that would occur at rated power.

. 5. 4. 2 Interlocking so that the period protection and downscale meter protection are in operation during startup.

I

5. 4. 3 Low pressure in any of the scram accumu-lator tanks.

Followed by page number III-44a

III-44n 6.

Reactor Nuetron Monitor System An instrumentation system as shown in Figure III-14 is provided to monitor the neutron level of the reactor from a startup through full powe r.

The instrumentation covers a range of 9 decades in three phases:

(1) Startup (2) Period or Intermediate Range (3) i Flux Level or Power Range.

6.1.. Startup Range With the initial fuel loading, a neutron source is inserted in the reactor core to assure a count rate of several neutrons per second. The startup instrumenta tion covers 7

the range upward to about 10 counts per minute.

Two channels of instrumentation monitor this phase of the ope ration. The primary neutron detectors are propor-tional counters. The average rate of the series of pulses are measured on a count rate meter with a log-arithmic scale in order to encompass six decades of measurement. The count rate is recorded continuously.

An additional circuit differentiates the log count rate and indicates the reactor period at this low level on a period mete r.

The approach of a short period at this low level is annunciated, h.2 Period Range This phase of the instrumentation system is concerned with the rate at which the neutron flux is increasing. Two channels of instrumentation monitor and control the reactor Followed by page number III 45

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III-56 Full flow of water is provided by one of two electric driven screen wash pumps, or during an outage of the electric pump, a stand-by diesel driven pump of full capacity. Normally, pressure is maintained in the fire system by a small fire system jockey pump and accumulator system. The pumps take suction from the circulating water intake structure.

Hose houses, hose racks, automatic sprinkler heads, and manual fire extinguishers are located throughout the plant.

A GO2 cystem is provided for controlling oil and electrical fires and for purging H2 fr m the generator. The CO2 is distributed to such points as lube oil tanks and lube oil filter s.

It is released automatically on temperature rise above a safe point.

12.2 Sanitary Service The sanitary system collects wastes from the plant build-ings, except those which are normally or potentially con-taminated with radioactive materials, and convey them by gravity into a septic tank.

A laundry consisting of an automatic washer and dryer is provided for washing any articles of protective clothing that are used during operation and maintenance in the controlled access areas of the plant. Water from this laundry is collected in a laundry waste tank and monitored before release to the circulating water discharge. In ad-dition to the laundry, there is a decontamination room containing a lavatory and shower. The discharge from these is also connected to laundry waste tank to allow monitoring before discharge.

12. 3 Potable Water Water from the lake is supplied to the domestic water system and maintained at approximately 50 psig by the fire system jockey pump, a pressure reducing valve, and a domestic water accumulator. The water is treated to comply with applicable health codes, after which it is distributed to all plant buildings.

12. 4 Heating and Ventilating I

Heating steam is supplied by an oil fired package boiler.

.{

III-57 A low pressure steam distribution ad condensate return system runs to all plant ouildings where they connect to:

Space h aaters for tha reactor building.

Air conditioners in the control room and laboratory.

Heating systems to all other building spaces such as office, shop, store room and locker room.

Unit heaters in controlled access areas.

The reactor containment vessel heating and ventilating system consists of units placed in advantageous locations and an exhaust fan discharging air into the vent stack.

During the summer, the units furni'sh cooled air by circulating lake water through the HV units' coils. A take-off from the service water pump discharge is pro-vided for this purpose. Outside air flows through a common air duct to all four units. Dampers at each HV unit regulate the amount of outside air..

The common air duct has a damper at the reactor containment vessel which closes automatically on power failure or when units are shut down.

Manual change-over to winter operation permits hot water to circulate through the same HV unit coils. This hot water is obtained by circulating water through a steam heat exchanger.

13. Electrical System

13. 1 General The electrical system consists of the main and auxiliary power system s The main power system consists of the main generator, main transformer and a high voltage switchyard, which delivers power into the Consumers Power Co. t ransmission system. Figure III-18 is the electrical single line diagram.

The auxiliary power system is designed to provide high reliability of service power supply to the station includ-(' '

ing power supply during shutdown and in the event of an

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III-59 e

accident. The general philosophy of design of this system has been to arrange the sources of power, wherever possible, in such a manner that any one local accident would still permit operation of the vital loads.

This is done primarily for good continuity of plant opera-tion. Fail-safe design of safety devices prevents an accident in case of complete power loss.

13. 2 Auxiliary Power Sources Auxiliary power for normal operation is obtained from the main generator leads through the house transformer connected to the 2400 volt bus. Loss of power on this bus for more than a short period scrams the reactor by means of a time delay undervoltage relay which trips contacts in the scram circuits. A transformer connected to the power company's transmission system supplies power to the 2400 volt bus during startup and shutdown and is automatically transferred to the bus in the event of failure of the house transformer. Each of the two 480 volt buses obtains power from a separate transformer connected to the 2400 volt bus.

The reactor instrumentation, safety circuits and annun-ciators are fed from two 120 volt preferred a-c buses.

Each bus is supplied by a motor-generator set with an auxiliary gas engine drive. The gas engine automatically drives the generator in the event of motor system failure.

Additional plant instrumentation which is' normally fed from conventional power sources is automatically trans-ferred to the preferred a-c bus if the normal source fails.

A 125 volt d-c battery system furnishes power for normal switchgear control, turbine control, and various emer-gency functions.

14 Reactor Servicing Equipment In general, fuel handling is accomplished by manual guidance and visual observation of all fuel handling operations. Water is used as the basic shielding material except for the transfer of spent fuel from reactor to storage pool. A lead shielded transfer cask is used for this operation, in conjunction with the overhead building crane.

k

III-60 When refueling is required, the reactor is shut down, and g

the reactor vessel head removed. An extension tank is bolted to the reactor head flange. This provides the additional water shielding required when fuelis removed or rearranged in the reactor core. A refueling platform is placed on the extension from which the refueling operation is controlled and observed.

The platform has an opening in it which allows the transfer casks.

to pass through into the reactor water for the fuel removal operation. A winch is used working through the same opening for the fuel shuffling operation. The platform is manually ro-tatable to enable vertical lifts with both transfer cask and shuf-fling winch.

To remove fuel from the reactor, the platform is rotated to en-able a vertical lift. The transfer cask is lowered into the ex-tension tank to a depth sufficient to provide water t ubmergance when the fuel is hoisted inside. The cask lower door is opened and a fuel grapple is lowered by cable into the reactor. An operator, using a long actuator pole, attaches the grapple to the fuel assembly to be removed. The fuel is hoisted out of the core guided by the operator. When clear of the core the actuator pole is removed and the fuel is hoisted into the cask. The lower door is closed and sealed and the cask is hoisted and moved to the fuel 1-storage pool. At the storage pool the cask is positioned over a storage rack and lowered into the water. The cask door is opened and the fuel lowered into the rack, guided manually by an operator using an actuator pole. The fuel grapple is removed from the fuel and engaged with a new fuel assembly. The new fuel, which may have an irradiated channel, is placed in the reactor by similar procedure.

While the cask is in transit to the storage pool, fuel can be re-arranged in the reactor core by using the fuel shuffling winch' mounted on the refueling platform. A fuel grapple is lowered by cable into the core. It is guided by an operator using an actuator pole. The fuel is grappled and hoisted clear of the core, the platform rotated, and the fuel replaced into a new location in the Core.

The transfer cask is also used for removal of control rods. The process is similar to fuel removal except that fuel must be re-moved to provide access for control rod removal.

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