ML20030A341

Chapter 5 to Final Hazards Summary Rept for Big Rock Poing, Reactor Sys Equipment

ML20030A341
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	11/14/1961
From:	CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
References
NUDOCS 8101090346
Download: ML20030A341 (22)

Text

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SECTION 5 Rev 2 (3/1 ?/62)

REACTOR SYSTEM EQUIPMENT 5.~ 1 GENERAL DESCRIPTION OF EQUIPMENT 5.1.1 The nuclear steam supply system is composed of the equip-ment in the main recirculation loop, plus those auxiliary systems required to provide a safe and operable system.

The main steam and condensate systei.- is shown sche-matica11y by Drawing M-106 with leganas identified by Drawing M-105.

5.1.2 The main recirculation loop consists of the rea.:or vessel and internals, including the core, control rods, and flow baffle; the steam drum, reactor recirculating pumpr., the interconnecting piping and valves, and the safety valves.

Piping and instrument diagram details are shown by Drawing M-121.

5.1.3 The auxiliary systems are: the shutdown cooling system, emergency cooling system, reactor cleanup demineralizer system, and the liquid poison system, all of which are shown on Drawing M-107.

5.1.4 The reactor vessel has been previously described in Section 4.1, and the liquid poison system in Section 4.5.

5.1.5 Data on pumps and heat exchangers are shown on Pages 18 and 19.

5.1.6 The design pressures and temperatures for the reactor system equipment are summarized below:

Equipment Design Pressure (psia) Design Temperatur e ( F)

Reactor Vessel 1715 650 Drum 1700 650 Piping 1625 600 Pump 1750 617 Valves 1765 615 5.2

' STEAM DRUM 5.2.1 The steam drum, with its piping, is mounted high up inside the enclosur e to perform the following functions:

5.2.1.1 Separate the steam from the steam-water mixture gener ated in the reactor core. The design criteria call for drum exit steam quality of 99.9 %.

i 5.2.1.2 Provide water storage to accommodate sur ges of water level and pressure between the reactor vessel and the drum.

To 'oR03de

4 S ction 5 Page I A Rev 2 (3/19/62)

Approximately 500 cubic feet of water are stored in the

' dr um. If, at full load operation, all steam voids in the core, reactor vessel and riser piping collapsed, the water storage available is sufficient to keep the downcomer piping inlets covered. Operationel transients and pump vibration will not occur as a result of steam drawn into the pump suction,

. and the supply of reactor recirculating water will be maintained.

5. 2.1. 3 -

~ Assure net positive suction head for the recirculating pumps to meet their design requirements. Drum water level is 65 feet above the center line of the pump suctions. The static head is sufficient to maintain flow during normal operation without pump cavitation; during transient conditions limited pump cavitation may occur.

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Secticn 5 Page '2 5.2.1.4 Provide natural circulation driving head to maintain flow in case the recirculating pumps are inoperative. It has been calculated that it will be possible to run at over 50% load on natural circulation alone with both pumps inoperative but free to rotate.

5.2.2 The drum has an over-all length of about 40 feet with an inside diameter of 78 inches. Design wall thickness is 4-3/8 inches minimum. Base material for shell and heads is SA-212-B, Fire Box. carbon steel clad with 5/32-inch minimum weld deposited type 309 and type 308 stainless steel on all internal surfaces. The cladding thickness is not

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considered in wall thickness calculations. Nozzles 4 inches and over are.SA-105-Gr. II carbon steel forgings clad internally with stainless steel. Nozzles under 4 f

inches are solid inconel SB-166. The drum internals are essentially stainless steel and inconel plate and strip, il

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,g A-167 and A-276 type 304 and SB-166 and SB-168.

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5. 2. 3 Design and construction of the drum satisfies the require-ments of the ASME Boiler and Pressure Vessel. Code,-

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,b Section I, and Code Cases 1270N and.1273N which include m. fL ~

,p g i-a hydrostatic test at 1. 5 design pressure. -Stainless cladding -

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. conforms to Section VIII, Part UCIsof the same Code. The drum is Code stamped. Design calculations for the drum have been made to cover the following:

5.2.3.1 ASME Code allowable stresses.

5.2.3.2 A detailed structural analysis of the shell nozzles and attach-ments to account for principal stresses and their combination for normal and transient power operation.

5.2.3.3 A transient analysis that concerns itself with the fatigue limits of the design.

5.2.4 The drum is supported from the concrete overhead structure by 8 constant support hangers. Movement of the drum due to thermal expansion of the piping and reactor vessel is compen-sated for by a specially designed support system which makes the downcomers, anchored at the same elevation as the reactor vecsel, into thermal rams to move the drum up as the rising temperature expands the components. Sidewise movement is controlled by guide rods that make the drum move on a line between the center of the drum and the reactor vessel anchor point. Additional rods keep the drum from rotating or skewing.

The suspension and support system as designed for main-taining the position of the steam drum is capable of with-standing the forces developed by a riser or downcomer line break.

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5. 3 RECIRCULATING PUMPS 9

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S ction 5 Page 3 Rev 1 (3/19/62) 5.3.1 Recirculation water from the steam drum is returned to'the reactor vessel by two vertically mounted recirculating pumps normally operating in parallel. The pumps are single stage, centrifugal, double suction with an overhung impeller. A water-lubricated bearing is provided internally to reduce the overhang. The pumps are of a mechanical seal-limited t

leakage design, with a total estimated outflow of 16 gallons per hour maximum per pump to the waste system.

5.3.2 The 17,000 gpm volume flow rate from each pump is essen-tially constant with two operating. With one pump operating, flow increa

~ about 207. over normal with a corresponding increese in ort positive suction head (NPSH) required.

5.3.3 The pump driver is a vertical, drip-proof induction motor of a conventional design rated for 400 HP at 900 rpm at 2300 volts. An oil-lubricated Kingsbury thrust bearing in the motor will absorb all residual pump and motor thru.M.

5.3.4 The pump casing design and construction is in accordance with the ASME Boiler and Pressure Vessel Code,Section VIII, latest edition, and the applicable lah=st interpretations of the nuclear code cases. The pump body, cover, and impeller are stainless-steel castings solution heat-treated to ASTM-A-381-58T, in compliance with ASTM-E71-52 standards for Class 2.

5.3.5-Three mechanical seals operating in series are supplied with each unit. The first seal and the graphitar water-lubricated bearing operate at system pressure and are protected from the system temperature by a heat shield and a supply of lower temperature (125 F max) reactor water that is circulated (12-15 gpm) by a small shaft pump through an external cooler. The cavities between the first and second seal, and between the second and third seal, are cooled (125 F max) by reactor water drawn off (15 gph max) the pump suction, passed through a heat exchanger and reduced in pressure through an adjustable orifice. The pressure reduction establishes that each seal sees only one-third of the total reactor operating pressure across its faces, or 500 psi at 1500 psi operating pressure. However, each seal is designed to take total reactor pressure across its faces if, for any reason, the other two seals malfunction. An instru-mentation system (Drawing M-237) is supplied to monitor temperature and pressure at each seal and will warn of any abnorrnal condition. Each seal has undergone a 1000-hour test, and each complete pump has been loop tested for 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> under reactor design conditions before delivery to the plant site.

5.3.6 The pumps are welded into and are partially supported by the recirculation piping, but the principal supports are two con-stant stipport hangers with subsidiary spring hangers. A vibration dampener at each pump is also supplied. A switch is mounted on each pump to warn of excessive vibration.

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5. 4.

PRIMARY LOOP PIPING AND VALVES

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5.4.1 Piping

5. 4.1.1 The primary loop piping connects all the major components

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of tha nuclear steam supply system. Six 14" risers carry the steam-water mixture from the reactor vessel to the steam

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drum. After the steam is separated, the recirculating water flows out of the drum down four 17" downcomers. The down-comers, in two groups of two each, join together into two

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24" pump suction headers. The two pump discharges are each 20" and return to the bottom of the reactor vessel.

5.4.1.2 Piping design criteria and material selection are in ccnfor-('

mance with the Code for Pressure Piping, ASA-B31.1-1955, latest revision and nuclear code cases N-1, N-7, 's-9 and

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N-10; as specified by the laws of the State of Michigan.

Fabrication and inspection of the pipe conforms to the appli-cable sections of Section I of the AS ME Boiler and Pressure Vessel Code. Piping suppcrts are designed in accordance with Section VI of the Code for Pressure Piping, ASA-B31.1-1955 latest revision.

I 5.4.1.3 The risers are stainless steel ASTM-A-358 type 316 seam welded pipe, which includes full radiograph and fluid pene-trant test of the seam. The remainder of the pipe is stain-less steel centrifugally cast ASTM-A-351 grade CF8M and meeting the requirements of nuclear code case N-9.

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fittings are stainless steel castings ASTM-A-351 grade CF8M and meeting the requirements of nuclear code case N-10.

Miscellaneous small piping satisfies to ASTM-A-376

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type 316. All shop welds and field welds are fully radio-graphed and fluid penetrant tested. Those sections of pipe which have been hot formed during fabrication have been solution

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heat treated and quenched. After the piping system has been erected it will undergo a hydrostatic test to

1. 5 times design pressure.

5. 4.1. 4 For the proper design of the piping system, a complete stress analysis was performed that took into account stresses due to I

expansion of the pipe and its connecting equipment, plus ex-ternally applied loads due to hangers, supports, anchors and vibration eliminators. Wall thickness design includes stresses f

due to normal operation and stresses due to over-pressures under transient conditions.

5.4.1.5 The piping is hung from constant support hangers at selected points, and anchored in the vertical direction at the same elevation as the reactor vessel supports. This effectively

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allows the pipe and other system components to expand freely up or down from the anchor point with very small resultant thermal expansion stresses. A system of vibration dampen-f

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ers are fastened to the piping to reduce possible vibration f

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U Section 5 Page 5 stresses.. An open bypass between the suct 'as of the two pumps will keep water flowing in the four downcome cs even if one pump

. is out of service. This feature prevents tilting of the steam drum due to unequal thermal expansion in the two sets of downcomer piping.

5.4.2 Valve s.

5. 4. 2.1 Two 24" gate valves, one on each suction side, and two 20" gate valves, one on each discharge side of the two recirculating pumps, serve as stop valves. The pt..mps are of a limited leakage. design and the valves are needed to assure shutoff if pump leakage becomes excessive for any reason. These valves are remote motor operated by extension shafts and can be hand operated from the far side of the shielding wall if necessary. Each pump has a startup bypass around the discharge stop valve with a 5" motor operated gate valve in it for control. These six valves are nor-mally operated electrically from the control room, but local push-button stations are mounted with the starters outsi~de the shielding wall.

5.4.2.2 These valves meet and exceed the standard of the ASA code for Pressure Piping for wall thickness. The valve bodies are stain-less steel castings to ASThi-A351 Grade CF8hi. All castings are fully radiographed after heat treatment to comply with ASThi-E71 standards for Class 2 castings, and are liquid penetrant tested.

After assembly, the valves will be hydrostatically tested at 1. 5 times design pressure.

5.4.2.3 The joint between body and bonnet on the valves is designed for seal welding, if, for any reason, Jeakage from the joint becomes excessive. The steam seal is a stuffing box with two sets of packing with a leakoff to waste in between.

5.4.2.4 In addition to the above valves, two 20" butterfly valves, one on the discharge of each recirculating pump, are installed to regulate pump discharge during the Research and Development Program period. These butterfly valves are motor operated and can be made inoperative in the open position. They are normally operated electrically from the control room, but local push-i button stations are mounted with the starters outside the shield-ing wall. The closing limit switch on the electrical operators of these valves is set to prevent valve closure of more than a predetermined amount. In the event that both valves are closed I

beyond the preset limit, the reactor will be scrammed by inde-pendent limit switches. Design and construction of these valves 1

are to the same standards as the gate valves discussed above.

The valve bodies are cast stainless steel ASThi-A351, Grade CF8.

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5. 5 SAFETY RELIEF VALVES

5. 5.1 The six safety-valves are located on the main steam drum and'

- will be set so that'the allowable pressure of 1870 psia in the I.

nuclear' steam supply system will not be exceeded.

-Rupture discs are mounted in the discharge of all safety valves.

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These discs are set at.10 psig (with +3 psi tolerance). Drain 3

-lines upstream at the discs carry ofTminor leakage to the

. building sump. A high temperature alarm in the drain signals excessive seat leakage.

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5. 6 TURBINE AND BYPASS CONTROL SYSTEMS 1

5.6.1 The turbine is arranged for two modes of control:

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= Initial Pressure Regulation (Base Loaded Operation).

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Speed Control (House Unit Operation).

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5. 6. 2 Normally, the turbine-generator is base loaded, and to avoid swings with accompanying' transients being felt by the reactor, the steam line pressure is maintained at a constant value by -

the initial pressure regulator (IPR). This pressure regulator positions the turbine admission valves to maintain a constant steam line pressure without regard to the generator or the transmission line system loads., The speed governor is nor-i c

mally backed off above the IPR control band, but overrides the IPR on increasing speed to keep the unit below the high speed (emergency) trip.

3, 5.6.3 Upon a sudden load loss, protective relaying separates the unit from the line, automatically transfers control from the

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initial pressure regulator to the sp eed governor through a solenoid transfer and reset device, and at the same time

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a Section 5

_Page 7 repositions the governor outer bushing to a predetermined set point at about the value of the house load. This causes a rapid closure of the turbine admission valve with the tur-bine speed remaining below trip speed. The turbine levels

' out close to synchronous speed and the turbine bypass valve opens to dump excess steam to the condenser and maintain correct reactor pressure.

t-5.6.4 To avoid frequent operating upsets during electrical storms, the generator line breaker will be fitted with an automatic device to reclose the breaker after a few cycles. A time delay will be installed to prevent control transfer under these circumstances, f

5.6.5 After a control transfer has taken place, the unit will be returned to initial pressure regulator control by manual ope ration.

5. 6. 6 The turbine bypass valve limits transient pressure increases in the main steam line by opening in response to two independ-ent pressur9 sensors located near the turbine stop valve which are set slightly a bove the normal IPR setting. The bypass valve is hydraulically actuated and electronically controlled to provide the rapid response necessary for load-loss transients.

5.6.7 When the 138 kv transmission line breaker is tripped as a recult of a transmission line fault, the generator load f alb to plant 'acx-iliary level and the turbine starts to apced up. Upon reaching the i

speed limit, the turbine admission valve starts to close under the influence of the turbine speed governor. As the admission valve closes, the pressure in the main steam line starts to rise and increases rapidly if corrective action is not taken in time. Since there is a time delay for the speed governor to act (i. e.. preventing control transfer per item 5. 6. 4) and hence for the admission valve to close, the rate of pressure change may be sufficiently rapid as to scram the reactor.

To enable the system to handle the generator load drop from full to auxiliary level without scramming the reactor, a signal is obtained from the transmission line relay opera-tion after an unsuccessful reclosure of the breaker, that lifts the bypass valve plug sufficiently to overcome the initial opening force. Immediately after the valve is cracked open, the control is returned to the pressure controller.

5.6.8 A condenser vacuum control to override the control system and close the bypass valve if condenser pressure and temperature rise to a preset level, is also provided.

5. 6. 9 Some of the safety features incorporated in the bypass system i

are the accumulator, to provide hydraulic power after an electric power failure; duplicate hydraulic pumps; the 138 kv breaker anticipatory system; a transducer burnout monitor; and the system operation monitor. Loss of hydraulic power is annunciated in the control room. All of the controls for i

the bypass system are located in the main control room.

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5. 7 SHUTDOWN COOLING SYSTEM

5. 7.-1 The procedure following reactor shutdown is to cool the nuclear steam supply system to allow lowering the water level to below the reactor vessel flange in order that the head may be removed for refueling.

5.7.2 The rate of system cooling is limited to In0*F/hr. Drum water level is maintained and the reactor recirculating pumps are Ecpc running to allow uniform cooling of the system.

5. 7. 3 The initial phase of cooling consists of bleeding steam into the main condenser via the main steam bypass valve.

(See Drawings M-106 and M-107).

The valve is throttled by a remote manual control to maintain the proper cooling rate. Metal temperatures on the steam drum and reactor vessel, as well as system water temperature and pressure, are recorded in the cort rol room for operation information.

5.7.4 When system pressure has dropped to 150 psig, the turbine seals become ineffective and it is necessary to close the bypae.s valve. The cooling is continued by recirculating reactor water through the shutdown cooling system. The skutdown system has a design pressure of 300 psig at 4c5'F, so that the changeover in the cooling operation from that through the bypass valve to'that through the shutdown coolers can be done at between 300 psig and 150 psig.

5.7.5 The shutdown cooling system consists of two vertical 500 gpm pumps and two heat exchangers (See Drawing M-107). No r --

mal power, and emergency power from the diesel generator, are available. Although a single system of one pump and one ex-changer will safely removeeore decay heat, two pumps and two exchangers are provided to assure system cooling reliability, and to provide for rapid cool-down of the system at the 100*F per hour rate to allow refueling to commence as soon as po s sible. The system is cooled to 212*F within about four hours after shutdown, at which time the vapor pressure in the reactor will be down to atmospheric and the reactor head bolts can be removed. Within eight hours after shutdown, the system is cooled to 120*F, and this temperature can be maintained by a single Inmp and exchanger.

5.7.6 As the shutdown system is rated at a lower pressuie than the reactor, it is isolated during normal operation by double block valves at both the inlet to the reactor vessel and return connections to the recirculaLion lines. A bleed-off orifice with a pressure switch between each pair of valves provides indication of leakage.

5.7.7 The block valves are motor operated with pressure inter-locks to prevent inadvertent opening while the reactor pres-sure is at 300 psig or higher.

. O Section 5 Page 9 5.7.8 When the shutdown system is inoperative, inhibited water from the cooling water system is routed through the shut-down system to avoid accelerated corrosion in stagnant line s.

5.7.9 Controls for both shutdown pumps and block valve control switches are located in the control room. In addition, a remote manual flow control valve is provided in the shut-down heat exchanger cooling water discha_rge which can be regulated to control the cooling rate. Thus, the entire cooling down operation can be carried out from the control 3

room, however, it may be necessary to locally valve off the fuel pit heat exchangers during initial shutdown to route more cooling water flow through the exchangers when both

- are in operation.

5. 8 EMERGENCY COOLING SYSTEM

5. 8.1 The emergency condenser is provided as a back up heat sink for reactor heat when the main condenser is unavailable.

5.8.2 The emergency condenser consists of a horizontal cylindrical tank with two internal sets of condensing tube bundles. It is located above the steam drum (See Drawing M-107). Eme r -

gency cooling is accomplished by condensing steam from the drum in the tube side of the condenser, which initially heats the stored water in the tank, and within about fifteen minutes, i

bulk boiling of the tank water commences.

5. 8. 3 Condensate flows from the tubes back to the drum, and flow through the tubes is maintained by the thermal siphon estab-4.

lished by the difference in density of the condensate in the return line and the steam i,n the supply line.

5.8.4 Steam generated in the emergency condenser shell is nor-mally non-radioactive, and is vented directly to atmosphere -

through a penetration in the top of the containment vessel.

5. 8. 5 There is no valve in the steam vent line, therefore, the inside of the eme rgency condenser shell is considered as an exten-sion of the outside of the containment vessel and is constructed in accordance with ASME Nuclear Code Cases 1270-N and 1272-N for an external pressure of 27 psig (corresponding to the design pressure of the containment vessel).

5.8.6 The emergency condenser is designed for a maximum capa-city of 32 x 10 Btu /hr or 4% of rated 240 Mwt reactor power.

This condition is predicated on the following considerations:

i 5.8.6.1 Loss of condenser vacuum initiates tripping the turbine stop valve and closure (arr prevents opening) of the main steam bypass valve.

5.8.6.2 As the steam pressure rises, the reactor scrams on either a high pressure signal or high flux signal.

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Section 5 Paja 10 e

's. 8. 6. 3 After scram, pressure continues to rice due c.2 the heat i,ane re.t e a in the core. Two duplicate pressure switches on ti.a rec.tr ere set aoove the hij'., grassure scram satiing to ~ initiate o o ra ti...

g of the emergency couacacer by eacr ining r..otor o arata-. v Ivec o

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on the two condensate return linco froc.a the canaencir.. Lusa ua 1:c.

5. 8. 6. 4 The motor operated valves open la o seconds and the syste:r. ic g

in full operation within 20 to 30 seconds. During this tirr.e, the pressure continues to rise for several seconds until the emergency condenser absorbs all of the decay heat generated and thereafter the ' pressure declines as the decay heat load falls off. The time lag and the heat transfer rate of the emergency condenser are selected to insure t ' the peak pressure rea'ched during this transient is substa.aially below the lowest setting; of the drum safety valve s.

5.8.7 After the decay heat load has fallen off, one of the condensate return valves is closed by a remote manual-switch in the control room to keep the cooling rate belo v 100*F/hr.

5. 8. 8 The water storage in the emergency condenser is sufficient for four hours operation without make up, thus, initial operation is dependent only upon DC power to operate the condensate valves.

After this period, if the main condenser has not been placed back into operation, make up is supplied to the tank by starting the demineralized water pump from the control room. A makeup valve at the condenser is opened and closed automatically by a level switch so that, once the demineralized water pump is started, makeup is automatic.

Ir. addition, a manual makeup connection is provided from the fire system in the event of failure of the demineralized water pump, 5.8.9 A radiation monitor on the emergency condenser vent stack will detect a release of primary steam, as may occur in the event of a tube rupture, and cause an alarm in the control room. Under this condition, the operator isolates the system by closing both condensate return valves and two motor operated valves on the steam inlet lines to the condensing tuaes. He can then reopen one system at a time to determine which tube bundle is still opera-tive. A single tube bundle is sufficient to remove reactor decay heat when the heat rate drops to 2% of 240 Mwt following shutdown.

5. 9 CORE SPRAY COOLING SYSTEM 5.9.1 The core spray cooling system is provided to prevent a core meltdown should the core become uncovered following an incident.

Initially the system is supplied from the same sources as the post-incident cooling system; i. e., from the fire protection system, followed by recirculation through the core spray pumps and heat exchanger (also see Section 3. 7). Cooling water is admitted into a circular sparger above the core which directs sprays onto the fuel elements.

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Section 5

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5.9.2 Cooling water is admitted to the spray sparger from the post-incident supply through the motor operated block valves which normally isolate the reactor from the tire protection system. - A check valve is also provided in this line for further insurance against back flow 1

of primary system water into the fire main. Opening of the ad-mission valves is actuated automatically by the simultaneous occurrence of a si.gnal from either of two low reactor water level t

switches along with permissive conditions as sensed by reactor

_ pressure switches. This is essentially an interlock which prevents admission valve opening while the reactor pressure is higher than t

fire protection system pressure. Also, a means for manual initiation of operation of the core spray system is provided (See Drawing M-123 ).

5.9.3 It is possible to supply the core spray system and post-incident spray system simultaneously. The core spray system is designed to supply sufficient water to remove core decay heat and minimize release of fission products into the containment atmosphere.

5.9.4 As indicated in Section 3. 7, which covers operation of the spray systems, when the water level in the containment vessel reaches elevation 587 feet the supply from the fire protecti.on system is shut off by a manual valve in the machine shop and one of the two 400 gpn vertical core spray pumps is started. These pumps are supplied from a suction header, consisting of screened drain openings in the bottom of the containment vessel and an external suction line which runs from under the containment vessel up to the pump room under the equipment loading dock. The pumps cannot operate until the water level has risen sufficiently to cover the suction line at about elevation 587 feet. The operating pump discharges into the core spray heat exchanger and back to the core spray supply header inside the containment vessel. Cooling. water is supplied to the heat exchanger from the fire p.rotection system by a remote man'ual valve which'is ope' rated from the control room.

5.9.5 A test tank and appropriate valving is provided in the core spray recirculation system so that the pump suction conditions and the flow characteristics of the system can be periodically tested. The test tank also serves as a chemical add tank for mixing rust inhibitor into the lower portion of the suction line.

3 5.10 REACTOR EMERGENCY ELECTRIC POWER 5.10.1 Certain of the reactor auxiliaries can be interruped for one-half to two hours without damage after a loss of the normal plant auxiliary power supply. These are shutdown cooling, reactor cooling water, and service water.

5.10. 2 Shttdown cooling is required in the event that power is lost during refueling (when the emergency condenser is inoperable). One reactor cooling water pump is required to remove heat from the fuel pit or shutdown coolers. Service water to the cooling water heat exchangers can be supplied through an alternate connection from the fire pro-tection system.

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5.10.3 Fuel pit cooling must be re-established within ten tu twenty i

hours to prevent overheating of the concrete surrounding the pool and eventual boiling in the pool.

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5.10.4 In the event of loss of normal auxiliary power, these systems

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can be supplied from the emergency diesel generator system (See Section 6. 7. 3). The emergency diesel generator also

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serves as a back-up power supply to the electric fire pump and core spray pumps in the event of loss of normal auxil-iary power.

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5. 11 REACTOR CLEANUP SYSTEM

5. 11. 1 The reactor cleanup system removes the corrosion products originating in the feedwater heaters, reactor vessel, recircu.

lating loop, piping and shutdown system equipment to maintain reactor water purity at or below 0.5 ppm dissolved solids with total solids at or below 2 ppm. Excess water removal from the primary system is also accomplished by the clean-up system.

5. 11. 2 Water from the reactor flows through the tubes of the regen-erative heat exchanger where it is cocica to approximately 210 degrees F by counter-current exchange with water being returned to the reactor from the cleanup demineralizer. The water then flows through a non-regenerative heat exchanger for further cooiing to 110 degrees F by exchange with reactor cooling water, and to the cleanup pump suction. The water is then pumped 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 degrees F.

A re-mote manual flow control vgive is supplied on the cleanup pump discharge to permit adjustment of the cleanup system flow rate.

5. 11. 3 As an alternate, reactor water can be made up with clean condensate while drawing off impure water to the radwaste demine ralize r.

5. 11. 4 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. Fresh resins are added by sluicing from the condensate demineralizer regener-ation facility into the cleanup demineralizer. However, it is possible to regenerate cleanup demineralizer resins during initial startup operations by routing them to the condensate demineralizer regeneration system. This would be done only until the resins become too radioactive to be handled i

in the regeneration system.

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5. 11. 5 The reactor cleanu system is shown in Drawing M-107.

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Section 5 Page 13 5.12 REACTOR SERV]CING EQUIPMENT 5.12.I Gene ral

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Most reactor servicing functions are carried cut with the 2se of a cab-controlled electric overhead semi-gantry crrne. The

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main trolley contains a 75-ton hook for handling the reacter head, fuel transfer cask and the fuel shipping cask. b addi-tien, a light outrigger heok (monorail crane) is previded for handling new fuel and for servicing auxiliary equ2pment.

5.12.2 Fuel Handling 5.12.2.1 In general, fuel handling is accomplished by manual guidance and visua; observatHn of all fuel handling operations. Water is used as the basic shielding material except for the trans-fer of spent fuel from reactor to storage pool. A lead shicided transfer cask is used for this oper ation, in cenjunction with the semi-gantry crane. In addition to the overhead c,emi-gantry crane, there are three other cranes or winches associated with fuel handline operation:, over the reactor.

There are the jib crane Dee paragraph 5.12. 2 3), transfer cask winch (see paragraph 5.12. 2. 2. 3 ), and the monorail crane. Means are pro" idea within all of these three hoists to restrict lowering speeds such that the maximum travel rate is no more than approximately 100 inches per minute.

5.12. 2. 2 Transfer Cask 5.12. 2. 2.1 The transfer cask is designed to accommodate fuel, channel-support tubes and control rods. The lead shielding thickness l

is 11" in the lower portion where needed for fuel element l

shielding and 6" in the upper portion whe re only the chanr.el-support tube needs shielding,. The total weight of the ca sk is approxima tely 24 tons.

5.12. 2. 2. 2 The imer cavity of the cask is approxima tely 12" in di.ameter.

Tlr upper end has a plag type door with a hole for a we.re rope pa ssage which aise serves to vent the ca sk.

The lower end has a door which can be opened to provide access to the full I

diamete r of the cavity. The door is operated by a single l

shaft running out the side near the top.

Turning the shaft swings the door and a jacking screw raises or lowers the door to facilitate a seal by compressing a rubber gasket.

This seal maintains the water level over the fuel. A h quid level ala rm annunciates when the level drops and a fil! con-nection is provided for adding make-up water if leakage should occur. In normal operation the cask is lowered into the water to the proper level so that filling or adding water is not necessary.

5.12.2.2.3 A motorized winch is provided on the side of the cask to hoist fuel or other assemblies. The hoist speeds are 24 ft/ min.

maximum, and 8 ft/ min. minimum. Gea red limit switches restrict the high speed to a specific elevation of rope travel and provide upper and lower limits of travel. An overload switch cuts power to the winch whenever a p*eset load is exceeded.

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Section 5 Page 14 5.12. 2. 2. 4 In operation, the transfer cask is supported from the over-head semi-gantry crane and lowered into reactor water through a rectangular opening in the refueling platform.

When loaded it is moved to the fuel storage pool where the unloading operation is also under water.

5.'12. 2. 3 Refueling Platform 5.12. 2. 3.1 The refueling platform is mace of aluminum grating supported oy structural aluminum channels and mounted on four grooved wheels. These wheels run on a machined flange of the exten-sion tank. The platform is manually rotated and is equipped with a foot-operated brake.

i 5.12. 2. 3. 2 A jib crane mounted on the platform is equipped with a winch which is identical to the one on the transfer cask. The upper limit of this winch is set such that it prevents hois' ting fuel too close to the surface. This winch also has an overload output to prevent damage to fuel or reactor internals in case of snagging while hoisting. The boom of the jib crane is man-ually pvsitionable, and, in con; unction with rotation of the platform, permits vertical hoisting over-all fuel positions in the core. Five travel speeds are provided with this jib crane, up ta.1 maximum speed of 24 feet per minute, with geared limit switches restricting speed to 8 feet per minute

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below the elevation of the top of the reactor core.

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5.12.2.3.3 The jib crane is used for re-positioning fuel in the core as l

required by the refueling schedule.

5.12. 2. 4 Refueling i

a 5.12. 2. 4.1 When rciueling is required, the reactor is shut down, cooled, and the reactor top shield plug and vessel head are removed.

A permanent extension tank, welded to the reactor below the flange is then filled with demineralized water. This provides the additional water shielding required when fuel is removed or rearranged in the reactor core. The refueling platform is placed on the extension tank from which the refueling operation is controlled and observed, i

5.12. 2. 4. 2 To remove fuel from the core, the refueling platform is rotated to a position allowing a vertical lift. The transfer cask is lowered into the extension tank to a depth sufficient to cover the fuel when the fuel is hoisted inside the cask. 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 bundle to be removed. The fuel is hoisted out of the core guided by t' - operator.

When clear of the core, the actuator pole is re.

ved and the fuel is hoisted into the cask. The lower d. ~

closed and sealed and the cask is hoisted and moved to t.

storage pool.

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i Secticn 5 Page 15 5.12.2.4.3 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 then removed from the fuel.

5.12. 2.4. 4 New fuel may be inserted into the core during the return trip with the transfer cask, or the monorail crane may be used in moving unirradiated fuel directly from the new fuel storage area into the core. Monorail crane lowering speeds are restricted to be consistent with the speeds of the other refueling cranes.

i 5.12.2.4.5 The reactor head has three access ports through which 12 fuel assemblies can be removed and replaced. Four fuel assemblics under each port are vertically in line i

with the three ports. By filling the extension tank with I

water and grappling fuel through these ports, the transfer cask can be used to remove these twelve assemblies.

5.12.3 Miscellaneous Reactor Tools l

5.12.3.1 Baffle Tool The baffle is hinged such that when four pie-shaped sections are opened, the full core is accessible from above. The baffle latches are first unlocked by a long pole with a hook on the end. The central handle of each section is then manually lifted and latched in the up or open position by using the same pole.

5.12.3. 2 Top Guide Grapple and Floats

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To unfold the top guide to provide access to channels and control rods, i ndividual fl. oats are attached to each beam using the actuator grapple. The beam locks are then actuated and the beams individually lifted, using the top l

guide grapple. The floats function to prevent accidental movement of the beams while in the up position.

4 5.12,3.3 Channel Replacement The fuel grapple is used for removal or replacement of chan-nels, however, to attach it in place a longer pole is needed due to the bail location which is below the orifice position.

5.12.3. 4 Control Rod Replacement The additional tools required for control rod removal con-sist of two long pole grapples. One is used to unlatch the p

coupling to the control rod drive and the other to remove the control rod from the reactor. Hoisting and transfer is accomplished by use of the transfer cask.

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i Section 5 Page 16 5.12.3.5 In-Core Flux Detector Assembly Replacement A grapple is provided which attaches to the cable of the trans-fer cask and can be attached to the in-core flux detector guide tube by an actuator pole. The detector hangs out the bottom of the transfer cask in this operation since its length exceeds the cavity. The uetector assembly does not require cooling.

However, the radiation level is low enough to permit trans-ferring the assembly to the fuel storage pool without closing 1

i the bottom door completely.

j 5.12.4 Spent-Fuel Storage Ibol 5.12.4.1 The spent-fuel storage pit, located adjacent to the reactor, 4

is utilized for the following purposes:

4 a)

Storing used fuel elements in racks until shipment.

b)

Underwater inspection of certain irradiated fuel elements.

c)

Storage of reactor head during refueling in the event that it is too radioactive to allow dry storage.

d)

Underwater loading of spent-fuel shipping cask.

4 e)

Storing damaged fuel channels, control rods, in-core flux tubes and other highly radioactive equipment until di spos al.

5.12. 4. 2 The 29-foot water depth in,the spent-fuel storage pool is sufficient to provide adequate shielding over irradiated fuel while being lifted up and lowered into the fuel shipping cask.

i 5.12. 4. 3 Normally, the water in the spent-fuel storage pit is main-tained below a temperature of 120*F by means of either of two 250 gpm pumps and heat exchangers. Each pump and exchanger is capable of maintaining the above temperature while removing the decay heat from one-half of a fully-irradiated core. For greater heating loads, both pumps and exchangers will be operated, or in the case of outage, the temperature may rise to a maximum of 150*F.

5.12. 4. 4 In order to insure sufficient water clarity in the fuel pit for good visibility, a 180 gpm bypass filter is provided in the cooling system and the cooling piping is of aluminum to avoid scaling. In the event of contamination from ruptured fuel elements, the fuel pit water can be recirculated through the radwaste demineralizer.

l' 5.12. 4. 5 The fuel pit is traversed by a moving platform with an electric winch to allow loading the fuel cask and moving

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fuel elements into the disassembly area.

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I-Section' 5 Page 17 r

5.12.4.6 Space is provided in the fuel pool for the storage of dis-carded ion chambers, control rods and fuel channels until i

such time that they can be shipped from the site or stored in the solid waste vault.

5.12. 5 Fuel Storage Geometry Considerations t

Fuel storage geometries, analyzed by means of two dimen-sional, three group diffusion theory calculations, indicate that the fuel spacing in the storage areas (either the new or irradiated fuel storage areas) presents no criticality hazard with 3. 2 percent enriched fuel of the first core fuel design. The center-to-center spacings in the square array of 13 inches within the new fuel storage racks and 12 inches within the irradiated fuel storage racks are sufficient to maintain keff 40. 8 for an infinite array.

These calculations are based on assumptions of: (a) The arrays of fuel bundles are completely flooded with water of 68'F, (in the new fuel storage area this is not possible unless the sphere fills with water); (b) The array of fuel bundles is infinite in both directions consistent with centerline spacing and bundles are infinitely long; and (c) The reactivity was that of unirradiated fuel in both the new fuel storage area and the spent-fuel storage pool. The spacing between fuel bundle holders and the configuration of structural beams and special blocks preclude insertion of additional bundles of the first core design in the open spaces in the array.

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cn REACTOR SYSTEM EQUIPMENT - PUMPS Data Sheet

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p Design Pressure Capacity Diff Service T ype No.

Psia Gpm Head-Ft Reactor Recirculating Ve rt-Cent.

2 1,750 16,000 76 s

Reactor Shutdown Ve rt-Cent.

2 315 500 75 Core Spray Ve rt-Cent.

2 140 400 324 Reactor Water Cleanup Hor-Canned 1

1.715 95 130 Fuel Pit Hor-Gent.

2 140 250 110 Control Rod Drive Triplex-Recip 2

2,815 25 1.975 Psig s 'n N E2 us

Cn REACTOR SYSTEM EQUIPMENT - HEAT EXCHANGERS g

Data Sheet

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p w

Shell Design Tube Design Duty Temp F Temp F Service Btu / Hr Flow Lb/Hr Psia CF In & Out Flow Lb/Hr Psia F

In & Out 6

Reactor Shutdown (2) 20 x 10 250,000 90 320 85 - 115 250,000 315 425 190 - 160 Emergency Condenser Operating 15,700 42 250 100 Initial 26,900 1,715 650 605 15 x 10 6 (External) 215 Final Transient 26,900 601 32 x 106 6

Core Spray 8 x 10 145,500 165 235 50 - 120 197,000 225 235 140 - 100 6

Non-Regen Reactor

4. 75 x 10 200,000' 90 200 70 - 93.7 47,500 1,715 600 210 - 110 Cleanup Regen Reactor 2 x 10 47,500 1,715 600 110 - 522 47,500 1,715 600 595 - 210 Cleanup Fuel Pit Coolers 3 x 10 125,000 90 150 70 - 94 125,000 90 150 119 - 95 ChE O

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O SECTION 6 Rev 1 (3/12/62)

POWER SYSTEM EQUIPMENT 6.1 TURBINE-GENERATOR SYSTEM 6.1.1 The turbine is a 3600 rpm, tandem-compound, double l

flow, condensing unit direct connected to a hydrogen cooled generator which in turn is connected through a reduction gear to an air-cooled exciter. Three points of extraction for feedwater heating are provided.

l, 6.1.2 The turbine is rated at 54,500 kw at 1000 psig, o degrees final superheat and 3-1/2 ir.ches of mercury absolute exhaust pressure with 3 percent makeup allowance and three feedwater heaters in service. The turbine is capable of operating continuously at 1450 psig, O degrees final superheat and 2 inches Hg back pressure with a maximum expected output of 75,000 kw.

6.1.3 The 13,800 volt, wye-connected generator is rated 70.588 kva, 0.8" power factor, 0.80 shor t cir cuit ratio at 30 psig hydrogen cooling pressure. If the expected maximum output of the nuclear steam supply system is achieved, the generator nameplate rating will be increased to 88,235 kva, 0.85 power factor, 0.64 short circuit ratio.

6.1.4 Besides conventional design criteria, all modifications necessary due to use of saturated steam from a boiling water reactor are incorporated in the turbine design. Particular attention is given in the design of the machine to the elimina-tion 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 extraction point; in addition, two external moisture separators are provided in the cross-over between the high-pressure and low-pressure sections.

l Materials used in the construction of the turbine are selected to minimize the wear caused by wet, oxygenated steam.

6.1.5 The flow paths through the turbine are shown in Drawing M-106, which also shows the extraction drains and v:nts.

Heat balance for system loading of 50,000 kw is illustrated by Drawing M-ll2.

6.1.6 Data on pumps and heat exchangers are shown on Pages 9 and 10.

6.2 MAIN TURBINE CONDENSER 6.2.1 The main turbine condenser is designed to perform the follow-ing functions: (a) Condense the steam exhausted from the turbine to obtain the desired heat uillization and vacuum; (b)

De-aerate the condensate and water from heater drains and other returns; (c) Serve as a heat sink for excess reactor steam dumped through the turbine bypass valve; and (d)

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

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ta POWER SYSTEM EQUIPMENT - PUMPS Data Sheet

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f Design Pressure Capacity Diff l

Service Type No.

Psia Gprn Head-Ft Condensate Ve rt-Cent.

2 265 1,000 520 i

Reactor Feed Hor-Cent.

2 2,015 (Hydro)

I,600 3,400 Service Water Ve rt-Cent.

2 2,100 88 Reactor Cooling Water Vert-Cent.

2 1,500 100 l

Condenser Circulating Ve rt-Cent.

2 24,500 25 Jockey Fire Vert-Cent.,

1 25 250

]

Electric Fire Ve rt-Cent.

I 1,000 254 Diesel Fire Vert-Cent.

I 1,000 254

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