ML20153G132

Proposed Tech Specs,Clarifying & Correcting Description of Electrical Power Sys

ML20153G132
Person / Time
Site:	La Crosse File:Dairyland Power Cooperative icon.png
Issue date:	02/21/1986
From:	DAIRYLAND POWER COOPERATIVE
To:
Shared Package
ML20153G064	List: ML20153G058 ML20153G132
References
NUDOCS 8602280031
Download: ML20153G132 (8)

Text

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P 2.5.3 Turbine Main Steam Bypass Valve The turbine main steam bypass valve shall be capable of bypassing at least 100 percent of rated steam flow directly to the condenser. - It shall be hydraulically operated and shall have a maximum full stroke operating time of.

1.5 sec. It shall be capable of automatic'or remote manual operation.

2.5.4 . Main Turbine Condenser System-2.5.4.1 The main turbine condenser system shall have sufficient capacity to condense full rated steam flow of the reactor while the turbine is being bypassed.

2.5.4.2 Loss of vacuum shall cause safety actions to be initiated as ,

specified in Table 1, or as a result of turbine trip.

2.5.5 Condensata Demineralization System The condensate deafneralization system shall be capable of removing soluble and insoluble impurities from the turbine condensate by ion exchange and filtration.

2 5.6 Feedwater System 2.5.6.1 The feedwater system shall be capable of returning condensate through a high pressure feedwater heater to:the forced circulation pump suction header by either of two centrifugal pumps.

2.5.6.2 The pumps shall be driven by constant speed motors through variable speed fluid couplings and shall be capable of either manual control or automatic control of a three-element feedwater control system.

2.5.o.3 The feedwater line shall have a check valve within the reactor building and a remote manually-controlled shutof f valve within the turbine building.

2.6 Deleted 2.7 REACTOR CORE AND VESSEL INTERNALS 2.7.1 Vessel Internals and Core Structure 2.7.1.1 - The vessel internals shall include a thermal shield, a core support skirt, a plenum separator plate, a bottom grid assembly, internal steam separators, a thermal shock shield, a baffle plate structure with a peripheral lip, steam dryers with support structure, emergency core spray tube bundle 'l, structure combined with fuel hold-down mechanism, and the reactor core structure.

WPl.6.8 8602290031 860221 PDR ADOCM 05000409 P PDR

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4.2.1.9 The containment building shall be isolated whenever the spent fuel storage well contains irradiated fuel which has decayed less than 43* days after exposure in a critical reactor and a shipping cask for irrradiated fuel is being moved by the crane on the 701 foot level or located within one cask length of the top of the spent fuel storage well or is within the spent fuel storage well. During cask movement near or at the FESW the water level in the FESW must be at least 16 feet above the top of the fuel storage rack (no more than 7 feet below the top of the FESW).

4.2.2 Reactor Vessel, Coolant, and Auxiliary Syscems 4.2.2.1 Additional penetrations to the systems containing reactor coolant shall be designed, manufactured, and tested according to the provisions of the ASME Boiler and Pressure Vessel Code and the ASA Code for Pressure Piping applicable as of June 1962. These additional penetrations shall be limited to instrument connections and piping connections, the latter being no larger than 1-in. inside diamater.

4.2.2.2 The reactor coolant shall be light water and shall conform to the following requirements.

CONDITION 1 Normal Limit Maximum Limit Chloride concentration .2 ppm .5 ppm pH 5.3 - 8.6 NA Conductivity 3 umho/cm 10 paho/cm The time spent above 3 umho/cm at 70*F -80*F and .2 ppm chloride should not exceed 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> per incident nor 2 weeks per year. If either time limit is exceeded an orderly shutdown shall be initiated within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> unless returned to within the limits. When the maximum conductivity or chloride limits are exceeded an orderly shutdown should be-initiated immediately. If the pH is outside the limits for a period of greater than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> an orderly shutdown shall be initiated.

CONDITION 2 & 3 Normal Limit Maximum Limit Chloride concentration .1 ppm NA pH _

5.3 - 8.6 NA vp Conductivity 5 paho/cm NA

• 43 days for off loading less than one half of the cora, i.e. less than 36 fuel elements. 51 days for off loading more than 36 fuel elements.

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n-REACTOR COOLANT SYSTEM BASES ,

4.2.2.5 SAFETY VALVES The safety valves are designed to meet the requirements of the ASME Boiler and Pressure Vessel Code. The reactor primary system overpressurization protection is sufficient to limit the pressure within the pressure-retaining boundaries to less than 1540 psig, which is less than 110% of the vessel design pressure of 1400 psig.

The safety valves have a ndnimum stamped relieving capacity of 294,612 lb. per hour at a relief pressure of 1390 psig and 302,160 lb per hour at a relief pressure of 1426 psig. Three safety valves are installed. The relieving capacity with one valve inoperable is sufficient to limit the primary system t pressure to less than 110% of the vessel design pressure during an abnormal transient with the highest pressure, which is the MSIV closure. A high pressure scram is initiated at 1325 psig and no credit is taken for the MSIV l closure scram signal, the power-flow scram, the overpower scram nor the pressure reduction due to automatic operation of the shutdown condenser heat sink.

During the postulated most limiting pressure transient, caused by a turbine trip without bypass valve operation with full scram on high power (120%),

reactor pressure would not reach 1390 psig, the lowest safety valve set point.

The safety valve function is therefore not expected to be required under the most limiting operational transient.

The testing frequency applicable to the safety valve function is provided to ensure operability and demonstrate reliability of the valves. The required testing interval varies with observed valve failures. Set point drif t within + 3% of the setpoint is not considered to be valve failure for the purposes of this test schedule. Setpoint drif t > + 3% of the setpoint will be cause to test additional valves in accordance with ASME Section XI test schedule. The popping setpoints are significantly below the 110% primary system design pressure safety limit. Therefore, adequate margin exists between the setpoint and the safety limit of 1540 psig. J For the purposes of establishing the test frequency, a valve shall be considered to have failed to function properly if the test relief pressure is determined to be outside of the allowable setpoint tolerance specified in the ASME Code to which the valve was constructed. For the LACBWR spring loaded valves which are constructed 'o the ASME Code Section VIII, 1962, and Nuclear Code Case N-1271, 1962, and wLich must be removed from the primary steam system to ccnduct the test, the allowable setpoint tolerance is + 3% of the set pressure. However, when the safety valve relief pressure is set prior to installing the valve on the reactor system, the maximum deviation of the test relief pressure from the specified set presure shall not exceed + 1.0% of the required set pressure.

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4.2.3 ELECTRICAL POWER SYSTEMS *~

TheOPERABILITYofthenormal-andbackupelectricalpowersystemsensures$ nit sufficient power will be availatie to . supply neces'ary. plant equipment during' normal operation and following a loss of offsite power event.

For normal operation, al1 power for-reactor and turbine auxiliary equipment'is

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supplied by the main generator through the unit auxiliary transformer or the reserve auxiliary tranpformer. During startup or shutdown, or in the. event of failure of power from the unit auxiliary transformer, power is supplied by a connection to 'the external grid via the reserve auxiliary transformer.

The onsite power distribution system is divided between two indepen' dent 2400 volt buses, and each component.of\ iuplicated equipment is supplied by a

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different bus. .The duplicated tquipment includes the primary system forced.

ci(culation pumps, the reactor feedtater pumps, the condensate pumps, and the condenser circulating water pumps. Each 2400 volt bus also supplies power to a 480 volt bus. If the supply to eitherc 480 volt bus fails, a connection may be manually made from the other 480 volt bus using a control switch in the control room. Duplicate equipment is also supplied by the 480 volt buses.

Each 480 volt bus normally supplies power to its associated essential-bus.-

The 480 volt Bus 1A supplies the 480 volt Essential Switchgear-Bus-1A.- The 480 volt Essential Switchgear Bus 1A supplies high pressure core spray. pump 1A and Turbine Building Motor Control center 1A which supplies the Turbine Building 120 volt Bus which in turn supplies the Turbine Building 120 volt Regulated Bus. Diesel Generator 1A serves as the backup power- supply for the 480 volt Essential Switchgear Bus 1A.

Turbine Building Motor Control Center 1A supplies power for the Reactor Plant Battery Charger. The Reactor Plant Battery Charger carries the load and charges the batteries on the Reactor Plant 125 volt DC-Bus. The Reactor Plant 125 volt DC Bus normally supplies power for Static Inverter 1A which supplies-power to essential reactor instrumentation and control circuits connected to 120 volt AC Non-Interruptible Bus 1A. In case of Static Inverter 1A trouble, an electro-mechanical transfer switch located in Static Inverter 1A automatically transfers to its. alternate power supply which is Turbine Building 120 volt Regulated Bus.

The 480 volt Bus 1B supplies power to the 480 volt Turbine Building Motor Control Center ID. Turbine Building Motor Control Center ID is the normal supply for the Generator Plant 125 volt DC Bus through the Generator Plant Battery Charger. The Generator Plant 125 Volt DC Bus normally supplies power to a Generator Plant 125 Volt DC distribution panel which supplies power to essential reactor instrumentation and control circuits connected to 120 Volt AC Non-Interruptible Bus IC. The static switch of the Static Inverter 1C is capable of supplying power from Static Inverter 1C or its reserve 120 Volt AC power source. In case of Static Inverter 1C trouble, the static switch will 32r(6)

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. automatically transfer from the output of Static Inverter IC-to the. reserve transformer which is supplied from Turbine Building Motor Control Center IA.,

The Reactor Plant 125 Volt DC Bus is capable of providing power to the Generator Plant 125 Volt DC Bus through a normally open tie breaker.

The Diesel Building 480 volt Essential Switchgear Bus 1B supplies high pressure core spray pump 1B, seal injection pump 1B, Reactor Building Motor Control Center 1A and Diesel Building 480 volt Motor Control Center. Diesel Generator IB is the emergency power supply for the Diesel Building 480 volt Essential Switchgear Bus IB. The Diesel Building 480 volt Motor-Control Center supplies power for the Diesel Building Battery Charger.- The Diesel Building Battery Charger carries the load of and enarges the batteries on the Diesel Building 125 volt DC Bus.

The Diesel Building 125 Volt DC Bus is the normal power supply for Static Inverter IB which supplica power to essential reactor instrumentation and control circuits connected to 120 Volt AC Non-Interruptible Bus 1B. In case-of Static Inverter IB trouble, its static switch can automatically transfer from the output of Static Inverter 1B to the output of its reserve transformer

.which is supplied by the Diesel Guilding 480 volt Motor Control Center. The Turbine Building 120 volt Regulated Bus is the reserve feed for the 120 Volt AC Non-Interruptible Bus IB. The normally open reserve tie breaker is administrative 1y controlled as are other tie breakers.

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MINIMUM FREQUENCIES FOR TESTING, CALIBRATING, AND/OR CHECKING OF INSTRUMENTATION - (cont'd)

Channels Action Minimum Frequency

6. Intermediate Range Test (10-10 and' Prior to each reactor Channels 3 and 4) 10-5 amps; start up if test has not period *) been performed within 30 days.

/ Check Once per shift when in service.

' ,7. Wide Range and Power Check by heat Monthly when in service.

Range (Channels 5, balance 6, 7, and 8) \

Monthly when in service and A. Nuclear Instru- Test

• prior to each reactor me..tation & Auto- startup if test has not been matic Gain Control performed within 30 days.

Sub-System.

B. Nuclear Instru- Check Once per shift when in f' mentation & Auto- service.

matic Gain Control Sub-System.

C. Automatic Gain Calibration At each refueling outage.

Control Sub-System NOTE: Testing of the Nuclear Instrumentation and Automatic Gain Control Sub-System shall be done concurrently.

8. Full Scram Circuits Test for hot Once a month.

short by means of built-in 3 test switch.

9. Area Radiation Calibration At least once per 18 months.

Monitors Test Quarterly Check Daily 5-8 (Reference 7 of the submittal letter also affects this page.)

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- d GENERAL MANAGER l l l SAFETY REVIEW COMMITTEE l l

TECHNICAL SUPPORT l LACBWR OPERATIONS l CONTRACTORS '

PLANT SUPERINTENDENT REVIEW COMMITTEE l

I I I I 1 QUALITY ASSURANCE ADMINISTRATIVE HEALTH & SAFETY ON-SITE TECHNICAL SUPERVISOR ASSISTANT SUPERVISOR SUPPORT l ENGINEERS

• I I MD HEALTH TECHNICAL QA/QC TRAINING PHYSICS SUPPORT TECHNICIANS l STENOGRAPHERS l SPECIALIST TECHNICIANS

! SECURITY & SKYETY SUPERVISOR i

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l I MECHANICAL INSTRUMENT &

MAINTENANCE ASSISTANT MECHANICAL l OPERATIONS SUPERVISOR l ELECTRICAL SUPERVISOR SUPERVISOR - MAINTENANCE SUPERVISOR INSTRQ(EKr AND MECHANICAL ELECTRICAL MAINTENANCE ASSISTANT TO TECHNICIANS i

SHIFT OPERATIONS SUPERVISOR SUPERVISORS & TRAINING 4

l PLANT.0PERATORSl RELIEF SUPERVISOR FIRE PROTECTION NOTE:

• Mechanical Fngineer 4

is Designated Engineer DAIRYLAND POWER COOPERATIVE Responsible for Fire'

. LACBWR FACILITY ORGANIZATION Protection Program.

A Technical Support

, Engineer is designated FIGURE 6.2.2-1 Emergency Planning 2 Coordinator i