ML18354A428

From kanterella
Jump to navigation Jump to search
Superseded by Amendment #14
ML18354A428
Person / Time
Site: Palisades Entergy icon.png
Issue date: 12/20/2018
From:
Consumers Power Co
To:
Office of Nuclear Reactor Regulation
References
Download: ML18354A428 (15)
v • d • e


Text

  • t The use of the finite element computer program permitted an accurate estimate of the stress pattern at various locations of the structure.

The following material properties were used in the program for the various loading conditions:

Load Conditions E Foundation (Psi) 5.0 x 106 5.0 x 10 6

concrete' E Shell (Psi) 6 6 concrete' 2.7 x 10 5.5 x 10

\l concrete (Poisson's Ratio) 0.17 0.17

a. concrete (Coeff of Expansion) 0.5 x 10- 5 E . (Psi) 0.1 x 10 6 0.1 x 10 6

soi 1 E . (Psi) 30 x 10 6 30 x 10 6

1 in er f . (Psi) 34,ooo y 1 iner .

(For definition of Load Conditions, see Appendix B. )

The major benefit of the program is the capability to predict shears and moments due to internal restraint and the interaction of the foundation slab relative to the. soil. The structur.e is analyzed assuming an uncracked homogeneous material. This is conservative because the decreased relative stiffness of a cracked section would result in smaller secondary shears and moments.

In arriving at the above-tabulated values of E, the effect of creep is included by using the following equation for long-term loads such as thermal load, dead load and prestress:

ECS = E.

Cl.

(E./(E

1. S

+ E.))

1.

Where:

Ecs = sustained modulus of elasticity of concrete.

ECl.. = instantaneous modulus of elasticity of concrete.

E.

1.

= instantaneous strain, in/in per psi.

E s

= creep strain, in/in per psi.

5-8

e Making and Curing Cylinders in Lab ASTM C-192 Compressive Strength Tests C-39 CONCRETE DESIGN MIXES Concrete A55re5ates Design Cement Fly Ash ProEortions b~ Wei5ht Stren5th Sks/Yd Sks/Yd ~ 3/4" 1-1/2" _L Water 4000 Psi 4.13 1.37 1420 1960 234

@90 Days 3,94 1.32 1283 1024 1.091 228 3.74 1.26 1149 662 761 906 217 5000 Psi 5.32 0.94 1333 1918 259

@ 28 Days 5.10 0.90 1221 996 1095 250 Water Reducing Agent - Walter Flood and Co was engaged to perform the necessary strength and shrinkage tests of various water reducing agents to establish the particular _additive with the most desirable characteristics for this application.

On the basis of these tests, Pozzolith 8, Improved, manufactured by Master Builders, was selected.

5.1.8 CONTAINMENT TESTING Tests will be made to determine.the initial leak tight and structural integrity of the containment.

5.1.8.1 Leak Tight Integrity Tests The objectives of these tests are:

(a) To determine the initial integrated leak rate for comparison with the 0.2%/day by volume at 55 psig and 283° F specified as the maximum permissible.

(b) To determine the characteristic leak rate variation with pres-sure so as to allow retesting at pressures less than design

  • pressure.

(c) To institute a performance history summary for both local leak and integrated leak rate tests.

The guidelines established for the tests are:

(a) The methods and equipment used during the initial tests will be such that the~ can be used for subsequent retests, thus avoid-ing test result variations due to changes of the methods or equipment.

5-41

2. After final assembly and calibration, the transmitter housing will be leak checked using helium and a mass-spectrometer leak detecting device.

(c) Pressurizer Pressure Transmitters The pressure transmitters whose output signal is used to generate the low-pressure trip signal for reactor trip and safety injection system actuation will be subjected to test conditions at least as severe as those used for the safety injection valves and high-pressure*safety injection flow transmitters.

(d) The following additional equipment has been or will be environmental tested for a 24-hour duration at concurrent 55 psig, 283° F and 100% relative humidity:

1. Containment area radiation monitor for containment isolation.
2. Containment air cooler fan motor unit.
3. Containment air cooler high-capacity service water valve.
4. Auxiliary devices such as solenoid valves associated with correct operation of the above items.
5. Samples of control, power and instrument cable were tested in a gamma pool at DBA temperature conditions to a total dose of 4.0 x 107 R which is twice the total 40-year integrated dose plus DBA. Physical tests in accordance with IPCEA were performed and the insula-tion met all requirements. The control and power cables were then tested to DBA concurrent temperature, pressure and humidity conditions while carrying rated current.

All samples passed the electrical and physical tests in accordance with IPCEA.

6.1.4 DESIGN ANALYSIS Ability to meet the core protection criteria is assured by the following design features:

(a) A high-capacity passive system which requires no outside source and will supply large quantities of borated water to rapidly recover the core after a major loss-of-coolant accident up to a break of the largest primary coolant line.

(b) A pumping and water storage system with internal redundancy which will inject borated water to provide core protection 6-12e Rev 4/10/69

l0- 4%and above 15% power. High rate-of-change of power a1arms are initiated at 1.5 dpm over the operating range of 10- 8% to 125~ power by the two wide-range channels. In adclf~ion, manual or automatic rod withdrawal prohibits between ~10-~ and 15~ power from the two wide-range channels prevent all regulating rods from being withdrawn, but do not prevent insertion.

This is an anticipatory trip which is not required to protect the reactor since the primary trip is high power level trip.

7.2.3.2 High Power Level A reactor trip at high power level (neutron flux) is provided to shut down the reactor when the indicated neutron power approaches an unsafe value. The high power trip signals are initiated by two-out-of-four coincidence logic fr9m the four power range safety channels. During normal plant operation with all coolant pumps operating, reactor trips are initiated when the reactor power level exceeds a nominal value of ~107'{o of indicated full power; this trip level represents a reactor power of no greater than 112% of full power when instrument and calorimetric errors are taken into ac-count. Provisions have been made to select different trip points for various combinations of primary coolant pump operation as de-scribed in the low primary coolant flow trip section.

The power range channels are equipped with a range change switch to increase the indicated power by a factor of 10. By use of the range change switch, indicated power is increased to provide full-scale indication at 12.5% power. This action also decreases the overpower trip from 107% to 10.7% to provide overpower trip pro-tection during low power operation.

Pretrip alarms are initiated at 10.4% or 104% of indicated full power depending upon adjustment of the range switch. The pretrip alarm signals are initiated by bistable trip units from the same channels which provide the reactor trip signals. The pretrip alarms provide annunciation in addition to rod withdrawal pro-hibit signals.

7.2.3.3 Low Reactor Coolant Flow A reactor trip is provided to protect the core from a power to flow mismatch. Provisions are made in the reactor protective system to permit operation at reduced power if one or more coolant pumps are taken out of service. For this mode of operation, the low flow trip points and the overpower trip points are simultaneously changed by a manual switch equipped with channel separation to the allowable values for the selected pump condition, thus pro-viding a positive means of assuring that the more restrictive settings are used. The switching arrangement is shown on Figure 7-4. The switch settings are readily visible to the operator.

The flow measurement signals are provided by summing the output of the differential pressure transmitters to provide an indication 7-4

of total coolant flow through the reactor. A reactor trip is ini-tiated by two-out-of-four coincidence logic from either of the four independent measuring channels when the flow function falls below a preselected value.

Pretrip alarms are initiated if the coolant flow function approaches the minimum required for reactor operation at the corresponding power level. A key-operated bypass switch allows this trip to be bypassed for subcritical testing of control rod ~ive mechanisms.

The trip bypass is automatically reset above 10-~ power.

7.2.3.4 High Pressurizer Pressure A reactor trip for high pressurizer pressure is provided to prevent excessive blowdown of the primary coolant system by relief action through the pressurizer power-operated relief or safety valves.

The trip signals are provided by four narrow range independent pressure transducers measuring the pressurizer pressure. A typical channel diagram is shown on Figure 7-2.

A reactor trip is initiated by two-out-of-four coincidence logic from the four independent measuring channels if the pressurizer pressure exceeds 24oo psia. This signal also opens the power-operated relief valves.

Pretrip alarms are initiated if the pressurizer pressure exceeds 2350 psia.

7.2.3.5 Thermal Margin/Low-Pressure Trip A trip is initiated by a continuously computed func_tion of primary coolant pressure and thermal power to prevent reactor conditions

_from violating a minimum departure from nuclear boiling ratio (DNBR).

At constant coolant flow, the temperature rise in the reactor is a function of power so that the variable trip can be effected by the adjustment of a pressure trip set point with reactor inlet and out-let coolant temperatures. At partial flow conditions, the changes in coolant temperature are such that the low thermal margin protec-tion is continued with no change required in the pressure set point function. The variable pressure trip set point is computed by the function, PTrip = ATJiot - Brcold - C. At design conditions, the values of the constants are A = 57, B = 30, and C = 15,SoO. These values may be adjusted in the field to accommodate changes in the process conditions. With the function suitably adjusted, reactor operation at less than the l.3 DNBR is not permitted. A further restriction on operation is that a trip is initiated if the pres~

sure is reduced to 1750 psia. The trip signal is initiated by a two-out-of-four coincidence logic from four independent safety channels, and audible and v;..eual pretrip alarms are actuated to provide for annunciation on approach to reactor trip conditions.

A block diagram of the TM/LPT function is shown on Figure 7-5.

7-5

An abnormally high main steam flow from either steam generator will cause the secondary pressure to drop rapidly.

Four pressure transmitters on each steam generator actuate trip units which are connected in a two-out-of-four logic to initiate the reactor protective action if the steam generator pressure drops below a preselected value. Signals from two of the four indicating meter relays from either steam generator will trip the reactor and close the main steam isolation valves on both steam generators.

Pretrip alarms are also provided.

A key-operated bypass switch allows this trip to be bypassed. This trip bypass is automatically reset above a selected pressure level.

7.2.3.9 Manual Trip A manual reactor trip is provided to permit the operators to trip the reactor. Manual actuation of either of two reactor trip push-button switches in the main control room causes direct interruption of the a-c power to the d-c power supplies feeding the CRDM elec-tromagnetic clutches.

7.2.4 SIGNAL GENERATION Four instrument channels are used to generate the signals necessary to initiate the automatic reactor trip action except for the loss of load and high rate-of-change trips where two measuring channels are used. The signal cable routing and readout drawer locations are separated and isolated to provide channel independence.

(1) The high rate-of-change of neutron flux signals are generated by the two wide-range measuring channels (Figure 7-6A) which monitor the flux from source level to 125% of full power.

These channels receive signals from flux monitors in the bio-logical shield around the reactor.

(2) The high neutron flux signals are supplied by four linear flux measuring channels (Figure 7-6B) covering the flux range from 0.1% to 125% power. These channels receive signals from ion chambers which monitor the full length of the core and are located in the biological shield around the reactor..

(3) The pressure, flow, thermal margin and water level trips are each actuated from signals generated by separate sets of transmitters. The primary coolant pressure is measured in the pressurizer, flow is measured by monitoring the .pressure difference across the steam generators, thermal measurements are taken from the reactor inlet and outlet piping in each loop and combined with primary coolant pressure to determine thermal margin, and steam generator water level and pressure are monitored in each steam generator.

7-7

Where the trip is to be allowed only in selected power ranges, a power dependent signal is supplied to the trip modules. Below the selected power levels (15% of rated power or less), these sig-nals provide inhibit action to all of the trips except high neutron flux, low thermal margin, low reactor coolant flow and excess steam flow. The high power rate-of-change trip is inhibited below 10-ltt;b power and also above 15% power. Each neutron flux measurement chan-nel supplies the automatic inhibit signals to trip in the same channel. Therefore, channel separation is maintained.

The CRDM clutches are separated into two groups. The clutches in each group are supplied in parallel with low-voltage, d-c power by an ungrounded feed line. Two a-c to d-c converters supply each feed line so that one converter being cut off does not cause re-lease of the clutches. The converters on each side are each sup-plied by a line from a preferred a-c bus to assure a continued source of power. Each line passes through two interrupters (each actuated by a separate trip path) in series so that, although both a-c lines must be de-energized to release the clutches, there are two separate means of interrupting each line. This arrangement provides means for the testing of the protective system.

The functions of reactor protective system Relays Kl, K2, K3, and K4 shown on Figure 7-lB are listed in Table 7-1.

TABLE 7-1 REACTOR PROTECTIVE SYSTEM RELAYS Con- Con-tact Function tact Function*

Kl-1 Oscillograph K2-l ***Rod Rundown Kl-2 Spare K2-2 Rod Rundown Kl-3 *Tr:i,p Lockout K2-3 Trip Lockout Kl-4 Trip.Lockout K2-4 Trip Lockout Kl-5 Diesel Generator Start (1-1) K2-5 Diesel Generator Start(l-2)

Kl-6 Trips Turbine K2-6 Trips Turbine Kl-7 **Trip Reset Lockout K2-7 Trip Reset Lockout Kl-8 Annunciator K2-8 Spare K3-l Rod Rundown K4-l Spare K3-2 Rod Rundown K4-2 Spare K3-3 Trip Lockout K4-3 Trip Lockout K3-4 Trip Lockout K4-4 Trip Lockout K3-5 Diesel Generator Start (1-1) K4-5 Diesel Generator Start(l-2)

K3-6 Trips Turbine K4-6 Trips Turbine K3-7 Spare K4-7 Spare K3-8 Spare K4-8 Spare

  • Trip Lockout - The reactor trip reset push button must be depressed to permit reactor start-up following a trip.
      • Rod Rundown - The control rods receive a "rods in" signal following a reactor trip which causes any "stuck" rods to be driven to the bottom of the core.

7-9 Rev 4/10/69

(13) The d-c clutch power supply circuits operate ungrounded so

.e that single grounds have no effect. The clutches are supplied in two groups by separate pairs of power supplies to further reduce the possibility of clutches being improperly held. The clutch impedances and load requirements are such that the ap-plication of any other local available voltage will not prevent clutch release, eg, connection of the clutch supply circuit to the battery distribution.circuit would cause the distribution circuit fuse to blow due to excessive current drain. Connec-tion to a 115 volt a-c circuit would have similar effects.

Connection to available low-voltage d-c, such as the nuclear instrumentation power supplies, would have no effect since these power supplies have insufficient capacity to supply the load.

7.2.8 POWER SOURCES The power for the protective system is supplied from four separate independent preferred a-c buses. Each preferred bus is supplied from the battery system through an inverter to assure an uninter-rupted, transient-free source of power.

Each preferred bus also has provision for connection to an instru-ment a-c bus to permit servicing of the inverters.

The distribution circuits to the preferred buses are provided with circuit protective devices to assure that individual circuit faults are isolated.

PHYSICAL SEPARATION The location of sensors and.the connection of sensing lines to the process loop are effected to assure channel separation and preclude the possibility of single events negating a system action. The process transmitters lo.cated inside the containment, which *are re-quired for short-term operation following a DBA, are housed in cabinets for mechanical and environmental protection. Protection from excessive temperature and humidity conditions is provided during a DBA for a period of time sufficient to complete their assigned function. The routing of cables from these cabinets* is arranged so that the cables are separated from each other and from power cabling to reduce the danger of comm.on event failures. This includes separation at the containment penetration areas. In the control room, the four nuclear instrumentation and protective system trip channels are located in individual compartments.

Mechanical and thermal barriers between these compartments reduce the possibility of comm.on event *failure. Outputs from the compo-nents in this area to the control boards are buffered so that shorting, grounding, or the application of the highest avai~able local voltages does not cause channel malfunction.

7-12

with some reactor parameters at values which would-norma1ly cause a trip. For these special operations, zero power mode bypass switches may be used to bypass the low flow, low steam generator level, low steam generator pressure, and the low thermal 'IIJIJ.'rgin/

low-pressure tripiC'unctions.**: These bypasses are automatically removed above 10-*~ power.

Interlocks to prohibit regulating group withdrawal are provided to prevent the reactor from reaching undesirable conditions. These interlocks are summarized in Table 7-3.

~BLE 7-3 Manual Manual Automatic Individual Sequential Sequential Withdrawal Control Group Control _Group Control Prohibit Conditions Mode Mode Mode Tavg Deviation x Tavg - Tref Deviation x Pretrip Overpower x x x Rod Drop x High Start-Ur.Rate (Between 10-LJ.% and 15% Power) x *X x The part-length rods may be moved manually either individually or as a group. A selector switc~*prevents simultaneous manual move-ment of the part-length and any other rods. The part-length rods have upper and lower limits of travel. -

7.3.2.2 Primary Pressure Regulating Two independent pressure channels provide suppressed range (1500 to 2500 psia) signals for control of the pressurizer heaters and spray valves. The output of either controller may be manually.selected to perform the control function. DUring normal operation, a small group of heaters is proportionally controlled to maintain operating pres-sure. If the pressure falls below the proportional band, all of the heaters are energized. Above the normal operating range, the spray valves are proportionally opened.to increase the spray flow rate as pressure rises. A small continuous flow is maintained through the spray lines at all times to keep the pipes warm and reduce thermal shock as the control valves open.

    • Four bypass switches will be provided. Each bypass switch will remove all four trip functions from one of the four protective system channels.

7-19

7.4.2 SYSTEM DESCRIPTION 7.4.2.1 Process Instruln.entation The following process instruments are associated with the reactor protective, reactor control or primary plant controls:

Temperature - Temperature measurements are made with precision resistance temperature detectors (RTDs) which provide a signal to the remote temperature indicating control and safety devices.

The following is a brief description of each of the temperature measurement channels:

(a) Hot Leg Temperature Each primary hot leg contains five temperature measurement channels. Four of these channels provide a hot leg tempera-ture signal to the thermal margin/low-pressure trip set point computers. The other hot.leg temperature measurement channel provides a signal to the loop Tavg and 6T summing computers in the .reactor regulating system. The five hot leg tempera-

. tures are indicated on the control paneL A high-temperature.

alarm.from the one . control channel is provided to alert the operator to a high-temperature condition.

(b) Cold Leg Temperature F.a.ch primary cold leg contains three temperature measurement channels. Two of the channels provide a cold leg temperature signal to the thermal margin/low-pressure trip set point com-puters. These two channels also provide cold leg temperature

'indication on the *control panel. The other cold leg tempera~

ture measurement channel provides either a signal to the loop Tavg and 6T summing computers or a wide-range tempera-ture recorder.on.the main control panel. A switch is provided t'o select the cold leg temperature measurement channel for loop Tavg and 6T. The cold leg channel not selected for loop Tavg and 6T is automatically switched to the wide-range tem~

  • perature recorder.

(c) Loop Average Temperature Each loop is provided with an average temperature sUIDID.er. The Tavg summer receives inputs from the control channel hot leg temperature detector and the. selected cold leg temperature detector and provides an average temperature output to the reactor regulating system and to a recorder.

The temperature recorders are equipped with two pens. One pen records the average temperature and the other pen records the programmed reference temperature signal (Tref) corresponding to turbine load. (First-Stage Pressure) 7-28

The contact output of each trip unit is fed to a singl~ channel of the reactor protective system. Thus, with two wide-range logarithmic channels, a separate rate trip signal is fed t.~ Channels A, B, C and D of the reactor protective system. The <10-'+% of full power rate-of-change bypass is initiated by the wide-range channel. level sf..~aL The level signal is fed to two trip units** sr,~ to trip above 10-'+%.

  • Contacts from each trip unit open above 10-~ to remove the rate trip bypass and to remove the zero power manually actuated bypass associated with a single channel. The zero power manual actuated bypass allows control rod drop testing, or rod withdrawal for other

.tests during shutdown. The trips bypassed are low flow, low steam generator level, low steam generator pressure, thermal margin/~ow pressure. These trips are automatically reactivated above 10- %

full power.

The >15% fUll power rate-of-change trip bypass for a particular channel is initiated by a bistable trip unit in the power range safety channel. Above 15% full power, the bistable trip unit resets closing a contact in parallel with the rate trip contact associated with that channel (A, B, C or D). This method of rate trip bypass permits maximllin independence of rate trip channels.

The rate-of-change of power pretrip alarm utilizes a single bistable trip unit (containing two sets of relay contacts) in each wide-range logarithmic channel. Each set of contacts feeds an auxiliary bistable trip unit in one of the channels of reacto.r *protective system. The auxiliary trip unit in turn initiates the rod withdrawal prohibit signal and pretrip alarm. The signal to the auxiliary trip unit is bypassed below lo-4% and above 15% of full power to avoid spuriotl.s alarms and rod withdrawal prohibits.

Reset of the trip units operates as described for the start-up channels.

Power Range Safety Channels - The four power range channels measure flux linearly over the range of 1% to 125% of full power. The de-tector assembly- consists of two uncompensated ion chambers for each channel. One detector extends axially along the lower half of the core while the other, which is located directly above it, monitors flux from the .upper half of the core. The upper and lower sections have a total active length of 12 feet. The d~c current signal from each of-the ion chambers is fed directly to the control room drawer assembly without preamplification. Integral shielded cable is used within the region of high neutron and gamma flux.

The signal from each chamber (top and bottom) is fed to independent amplifiers. The output of the amplifiers is indicated, compared and summed. The individual amplifier output is indicated on the amplifier drawer. The outputs are compared with each other to in-dicate axial flux tilts. The summed output of the two amplifiers is indicated, recorded and compared with averaged summed outputs 7-35

TABLE 9-8A (Contd)

Position Valve Normal Shutdown After Loss No Valve Description Position Position of Air Service Water System (Figure 9-1 and Figure 9-2) 0823 Component Cool HX Dischg 0 0 0 0824 Return From Containment Coolers 0 0 0 0825 Eng Safe Room Cooler Supply c c 0 0826 Component Cool HX Dischg 0 0 0 0835 Turbine LO Cooler Stop Bypass c c 0 0836 Turbine LO Cooler Stop Bypass c c 0 0838 Normal Cont Cooler Control 0 0 c 0839 Generator H2 Cooler Stop Bypass c c 0 0843 Normal Cont Cooler Control 0 0 c 0844 Critical Service Wtr Header Iso 0 0 0 0845 Critical Service Wtr Header Iso 0 0 0 0846 Critical Service Wtr Header Cross-connect 0 0 0 0847 Supply to Containment Coolers 0 0 0 0852 Generator Exciter Cooler Supply Bypass c c 0 0857 Critical Service Wtr Header Cross-connect 0 0 0 0861 8" Return From Cont Coolers c c 0 0862 Containment Cooler Supply

  • 0 0 0 0863 Normal Cont Cooler Control 0 0 c 0864 8" Return From Cont Coolers c c 0 0865 Containment Cooler Supply 0 0 0 0867 8" Return From Cont Coolers c c 0 0869 Containment Cooler Supply 0 0 0 0870 Containment Cooler Supply 0 0 0 0872 Normal Cont Cooler Control 0 0 c 0873 8" Return From Cont Coolers c c 0 0876 Diesel Generator Cool Supply 0 0 0 0877 Diesel Generator Cool Supply 0 0 0 0879 Backup Cool Safeguards Pumps c c c 0880 Backup Cool Safeguards Pumps c c c 0884 Diesel Generator Cool Supply c c 0 0885 Diesel Generator Cool Supply c c 0 1318 Service Wtr Pump Header Isolation 0 0 0 1319 Service Wtr Pump Header Isolation 0 0 0 1359 Noncritical Service Wtr Header Isola-tion 0 0 c Component Cooling System (Figure 9-6) 0909 Letdown HX Return 0 0 0 0910 Component Cool to Cont Isolation 0 0 0 0911 Component Cool From Cont Isolation 0 0 0 0913 Supply Safeguards Pumps c 0 0 9-33c Rev 5/5/69

TABLE 9-8A (Contd)

Position Valve Normal Shutdown After Loss No Valve Description Position Position of Air 0915 Comp Cool Surge Tk Vent 0 0 c 0918 Comp Cool Surge Tk Makeup c c c 0937 Supply to Shutdown HX c 0 0 0938 Supply to Shutdown HX c 0 0 0940 Component Cool From Cont Isolation 0 0 0 0944 Supply to Spent Fuel HX 0 0 c 0945 Supply to Comp Cool HX 0 0 0 0946 Supply to Comp Cool HX 0 0 0 0947 Supply to Safeguards Pumps 0 0 0 0948 Supply to Safeguards Pumps 0 0 0 0949 Supply to Safeguards Pumps 0 0 0

. 0950 Return From Safeguards Pumps c 0 0 0951 Return From Safeguards Pumps c 0 0 Main Steam, Main and Auxiliary Turbine Systems (Figure 9-9) 0501 Main Steam Isolation Valve 0 0 As Is (Accumulator) 0510 Main Steam Isolation Valve 0 0 As Is (Accumulator) 0511 Steam Bypass Valve c Open for Bleed c (Note I) 0521 Steam to Aux Turbine Feed Pump c 0 0 Service and Instrument Air Systems (Figure 9-8)

Valve 1211, Instrument Air Supply to Containment, is open during reactor opera-tion or reactor shutdown and fails open on loss of air.

Process Sampling System (Figure 9-16)

Air-operated process sampling valves are normally closed unless sampling a specific point. All air-operated valves fail close .

Radioactive Waste Treatment System (Figures 11-2 and 11-3)

All air-operated valves in the radioactive waste treatment system, including liquid and gas discharge stop valves, fail close upon loss of instrument air.

Heating, Ventilation and Air Conditioning Reference Section 9.8.4.

Shield Cooling System During normal reactor operation and reactor shutdown, one of two air-operated shield cooling supply valves is open. Upon loss of instrument air, both supply valves fail open.

Note 1: A hand operator is provided on the steam bypass valve.

9-33d Rev 5/5/69

~---------------------------------------------~ ---**-

TABLE 9-13 DESIGN PARAMETERS Chemical and Volume Control System 1.1

  • General Normal Letdown Flow 40 Gpm Normal Purification Flow Rate 40 Gpm Normal Charging Flow 44 Gpm Primary Coolant Pump Controlled Bleed-Off (4 Pumps) 4 Gpm Normal Letdown Temperature at Loop 547.8° F Normal Charging Temperature at Loop 425° F Ion Exchanger Operating Temperature 120° F 1.2 Regenerative Heat Exchanger ~*E56 Quantity 1 Type Shell and Tube, Vertical Normal Heat Transfer 6 6.6 x 10 Btu/Hr Code ASME III , .Class A Shell Side (Letdown)

Fluid Primary Coolant, 1 Wt  %

Boric Acid, Maximum Design Pressure 2485 Psig Design Temperature 650° F Materials Stainless Steel Tube Side (Charging)

Fluid Primary Coolant, 6-1/4 Wt  %

Boric Acid, Maximum Des.ign Pressure 2735 Psig Design Temperature Materials Stainless Steel 9-65 Rev 4/10/69

TABLE 9-13 (Contd)

Operating Parameters - Regenerative Heat Exchanger Maximum Unbalanced Maximum Charging With Maximum Unbalanced Shell Side (Letdown) Normal Heat Transfer Purification Letdown

.Flow - *** Gpm 40 40 120 120 Inlet Temp - °F 547.8 547.8 547.8 547.8 Outlet Temp - °F 238 176 361 450 Tube Side (Charging)

Flow - Gpm 44 133 124 33 Inlet Temp - °F 120 120 120 120 Outlet Temp - °F 425 240 325 516 Heat Transfer - Btu/H.r 6.6 x 10 6 7.81 x 10 6 12.41 x 10 6

6.73 x 10 6

1.3 Letdown Orifice - R02003, R02004 and R02005 Quantity 3 Capacity, Each 40 Gpm Design Pressure 2485 Psig Design Temperature 550° F Normal Temperature of Fluid 238° F Maximum Temperature of Fluid 470° F Normal Downstream Pressure 470 Psig Normal Upstream Pressure 1525 Psig Material Stainless Steel Fluid Primary Coolant, 1 Wt  %

Boric Acid, Maximum 1.4 Letdown Heat Exchanger - E58 Quantity 1 Type Shell and Tube, Horizontal 9-66

Retrieved from "https://kanterella.com/w/index.php?title=ML18354A428&oldid=2397206"