Text

Proposed Tech Spec Upgrading Emergency Feedwater Sys to Comply W/Requirements of NUREG-0737,Item II.E.1.2
	ML17325B680
Person / Time
Site:	Arkansas Nuclear
Issue date:	09/12/1984
From:	; ARKANSAS POWER & LIGHT CO.
To:	;
Shared Package
ML17325B678	List: ML17325B677; ML17325B680;
References
	RTR-NUREG-0737, RTR-NUREG-737, TASK-2.E.1.2, TASK-TM NUDOCS 8409170086
	Download: ML17325B680 (42)
	v • d • e

3.4 STEAM AND POWER CONVERSION SYSTEM A~1i Applies to the turbine cycle components for removal of reactor decay heat.

~0b 'ective To specify minimum conditions of the turbine cycle equipment necessary to assure the capability to remove decay heat from the reactor core.

S ecifications 3.4. 1 The reactor shall not be heated above 280'F unless the following conditions are met:

1. Capability to remove decay heat by use of two steam generators.

2. Fourteen of the steam system safety valves are operable.

3. A minimum of 16.3 ft. (107,000 gallons) of water is available in the condensate storage tank.

4. Both EFW pumps and their flow paths are operable.

5. Both main steam block valves and both main feedwater isolation valves are operable.

3.4.2 Initiate functions of the EFIC system which are bypassed at cold shutdown conditions shall have the following minimum operability conditions:

a. "low steam generator pressure" initiate shall be operable when the main steam pressure exceeds 750 psig.

b. "loss of 4 RC pumps" initiate shall be operable when neutron flux exceeds lOX power.

c. "main feedwater pumps tripped" initiate shall be operable when neutron flux exceeds lOX power.

F 4.3 The automatic steam generator isolation system within EFIC shall be operable when main steam pressure is greater than 750 psig.

Amendment No. 50 40 8409170086 8409i2 PDR ADOCK 050003i3 P PDR

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3.4.4 Components required to be operable by Specification 3.4.1, 3.4.2, and 3. 4.3 shall not be removed from service for more than 24 consecutive hours. If the system is not restored to meet the requirements of Specification 3.4. 1, 3.4.2 and 3.4.3 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the reactor shall be placed in the hot shutdown condition within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . If the requirements of Specification 3.4. 1, 3.4.2, and 3.4.3 are not met within an additional 48 .hours, the reactor shall be placed in the cold shutdown condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

3.4. 5 If the condition specified in 3. 4. 1.4 cannot be met:

1. With one EFW flow path inoperable, the unit shall be brought to hot shutdown within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months , and if not restored to an operable status within the next 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months , the unit shall be brought to cold shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or at the maximum safe rate.

2. If both EFW trains are inoperable, restore one train to operable status within one hour or be in hot shutdown within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and cold shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or at the maximum safe rate.

3. If both EFM trains and the AFM pump are inoperable, the unit shall be immediately run back to <5X full power with feedwater supplied from the MFW pumps. As soon as an EFM train or the AFW train is operable, the unit shall be placed in cold shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or at the maximum safe rate.

Amendment No. 50 40a

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Bases The Emergency Feedwater (EFM) system is designed to provide flow sufficient to remove heat load equal to 3< percent full power operation. The system minimum flow requirement to the steam generator(s) is 500 gpm. This takes into account a single failure, pump recirculation flow, seal leakage and pump wear.

To support loss of main feedwater analyses, steam line/feedwater line break analyses, SBLOCA analyses, and NUREG-0737 requirements, the EFIC system is designed to automatically initiate EFW when:

l. all four RC pumps are tripped

2. both main feedwater pumps are tripped

3. the level of either steam generator is low

4. either steam generator pressure is low,

5. ESAS ECCS actuation (high RB pressure or low RCS pressure)

The EFIC system is also designed to isolate the affected steam generator on a steam line/feedwater line break and supply EFW to the intact generator according to the following logic:

If both SG's are above 600 psig, supply EFW to both SG's.

If one SG is below 600 psig, supply EFM to the other SG.

If both SG's are below 600 psig, but the pressure difference between the two SG's exceeds 150 psig, supply EFW only to the SG with the higher pressure.

If both SG's are below 600 psig and the pressure difference is less than 150 psig, supply EFM to both SG's.

b ~i At cold shutdown conditions all EFIC initiate and isolate functions are 11 <<

bypassed except low steam generator level initiate. The bypassed functions h 1 pi di i id Specification 3.4.2. "Loss of 4 RC pumps" initiate and "low steam generator if' i Amendment No. 50 41

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pressure" initiate are the only shutdown bypasses to be manually initiated during cooldown. If reset is not done manually, they will automatically reset. Main feedwater pump trip bypass is automatically removed above 10K power.

In the event of loss of main feedwater, feedwater is supplied by the emergency feedwater pumps, one which is powered from an operable emergency bus and one which is powered from an operable steam supply system. Both EFM pumps take suction from the condensate storage tank. Decay heat is removed from a steam generator by steam relief through the turbine bypass, atmospheric dump valves, or safety valves. Fourteen of the steam safety valves will relieve the necessary amount of steam for rated reactor power.

The minimum amount of water in the condensate storage tank would be adequate for about 4.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months of operation. This is based on the estimate of the average emergency flow to a steam generator being 390 gpm. This operation time with the volume of water specified would not be reached, since the decay heat removal system could be brought into operation within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or I less.

41a

3.5.1.7 The Decay Heat Removal System isolation valve closure setpoints shall be equal to or less than 340 psig for one valve and equal to or less than 400 psig for the second valve in the suction line.

The relief valve setting for the DHR system shall be equal to or less than 450 psig.

3.5.1.8 The degraded voltage monitoring relay settings shall be as follows:

a. The 4. 16 KV emergency bus undervoltage relay setpoints shall be > 3115 VAC but < 3177 VAC.

b. The 460 V emergency bus undervoltage relay setpoints shall be

>'423 VAC but < 431 VAC with a time delay setpoint of 8 seconds +1 second.

3.5.1.9 The following Reactor Trip circuitry shall be operable as indicated:

1. Reactor trip upon loss of Main Feedwater shall be operable (as determine/ by Specification 4. l.a, items 2 and 36 of Table 4.1-2) at greater than 5X reactor power. (Nay be bypassed up to 10X reactor power.)

2. Reactor trip upon Turbine Trip shall be operable (as determined by Specification 4. l.a, items 2 and 42) at greater than 5X reactor power. (May be bypassed up to 20X reactor power.)

3. If the requirements of Specifications 3.5.1.9. 1 or 3.5. 1.9.2 cannot be met, restore the inoperable trip within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or bring the plant to a hot shutdown condition.

3.5.1.10 The control room ventilation chlorine detection system instrumentation shall be operable and capable of actuating control room isolation and filtration systems, with alarm/trip setpoints adjusted to actuate at a chlorine concentration of < 5ppm.

3. 5. 1. 11 For on-line testing of the Emergency Feedwater Initiation and Control (EFIC) system channels during power operation only one channel shall be locked into "maintenance bypass" at any one time.

If one channel of the NI/RPS is in maintenance bypass, only the corresponding channel of EFIC may be bypassed.

Amendment No. 68, HX, 69 42a

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Bases Every reasonable effort will be made to maintain all safety instrumentation in operation. A startup is not permitted unless the requirements of Table 3.5. 1-1, Columns 3 and 4 are met.

Operation at rated power is permitted as long as the systems have at least the redundancy requirements of Column 4 (Table 3.5. 1-1). This is in agreement with redundancy and single failure criteria of IEEE 279 as described in FSAR, Section 7.

There are four reactor protection channels. Normal trip logic is two out of four. Required trip logic for the power range instrumentation channels is two out of three. Minimum trip logic on other instrumentation channels is one out of two.

The four reactor protection channels were provided with key operated bypass switches to allow on-line testing or maintenance on only one channel at a time during power operation. Each channel is provided with alarm and lights to indicate when that channel is bypassed. There will be one reactor protection system channel bypass switch key permitted in the control room.

Each reactor protection channel key operated shutdown bypass switch is provided with alarm and lights to indicate when the shutdown bypass switch is being used, The source range and intermediate range nuclear flux instrumentation scales overlap by one decade. This decade overlap will be achieved at 10- amps on the intermediate range scale.

The ESAS employs three independent and identical analog channels, which supply trip signals to two independent, identical digital subsystems. In order to actuate the safeguards systems, two out of three analog channels must trip. This will cause both digital subsystems to trip. Tripping of either digital subsystem .will actuate all safeguards systems associated with that digital subsystem.

Because only one digital subsystem is necessary to actuate the safeguards systems and these systems are capable of tripping even when they are being tested, a single failure in a digital subsystem cannot prevent protective action.

Removal of a module required for protection from a RPS channel will cause that channel to trip, unless that channel has been bypassed, so that only one channel of the other three must trip to cause a reactor trip. Thus, sufficient redundancy has been built into the system to cover this situation.

Removal of a module required for protective action, from an analog ESAS channel will cause that channel to trip, so that only one of the other two must trip to actuate the safeguards systems. Removal of a module required 43

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for protective action from a digital ESAS subsystem will not cause that subsystem to trip. The fact that a module has been removed will be continuously annunciated to the operator. The redundant digital subsystem is still sufficient to indicate complete ESAS action.

The testing schemes of the RPS, the ESAS, and the EFIC enables complete system testing while the reactor is operating. Each channel is capable of being tested independently so that operation of individual channels may be evaluated.

The EFIC is designed to allow testing during power operation. One channel may be placed in key locked "maintenance bypass" prior to testing. This will bypass only one channel of EFW initiate logic. An interlock feature prevents bypassing more than one channel at a time. In addition since the EFIC receives signals from the NI/RPS, the maintenance bypass from the NI/RPS is interlocked with the EFIC. If one channel of the NI/RPS is in maintenance bypass, only the corresponding channel. of EFIC may be bypassed.

The EFIC can be tested from its input terminals to the actuated device controllers. A test of the EFIC trip logic will actuate one of two relays in the controllers. Activation of both relays is required in order to actuate the controllers. The two relays are tested individually to prevent automatic actuation of the component. The EFIC trip logic is two (one-out-of-two).

Reactor trips on loss of all main feedwater and on turbine trips will sense the start of a loss of OTSG heat sink and actuate earlier than other trip signals. This early actuation will provide a lower peak RC pressure during the initial over pressurization following a loss of feedwater or turbine trip event. The LOFW trip may be bypassed up to 10K to allow sufficient margin for bringing the MFW pumps into use at approximately 7X. The Turbine Trip trip may be bypassed up to 20K to allow sufficient margin for bringing the turbine on line at approximately 15K.

The Automatic Closure and Isolation System (ACI) is designed to close the Decay Heat Removal System (DHRS) return line isolation valves when the Reactor Coolant System (RCS) pressure exceeds a selected fraction of the DHRS design pressure or when core flooding system isolation valves are opened. The ACI is designed to permit manual operation of the DHRS return line isolation valves when permissive conditions exist. In addition, the ACI is designed to disallow manual operation of the valves when permissive conditions do not exist.

Power is normally supplied to the control rod drive mechanisms from two separate parallel 480 volt sources. Redundant trip devices are employed in each of these sources. If any one of these trip devices fails in the untripped state, on-line repairs to the failed device, when practical, will be made and the remaining trip devices will be tested. Four hours is ample time to test the remaining trip devices and, in many cases, make on-line repairs'.

Amendment No. 50, 80, 61 43a

The Degraded Voltage Monitoring relay settings are based on the short term starting voltage protection as well as long term running voltage protection.

The 4. 16 KV undervoltage relay setpoints are based on the allowable starting voltage plus maximum system voltage drops to the motor terminals, which allows approximately 78K of motor rated voltage at the motor terminals. The 460V undervoltage relay setpoint is based on long term motor voltage requirements plus the maximum feedwater voltage drop allowance resulting in a 92K setting of motor rated voltage.

The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables during and following an accident. This capability is consistent with the recommendation of Regulatory Guide 1.97, "Instrumentation for Light-Water-Cooled Nuclear power Plants to Assess Plant Conditions During and Following an Accident", December 1975 and NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations".

The OPERABILITY of the chlorine detection system ensures that sufficient capability is available to promptly detect and initiate protective action in the event of an accidental chlorine release. This capability is required to protect control room personnel and is consistent with the recommendations of Regulatory Guide 1.95, "Protection of Nuclear Power Plant Control Room Operators against an Accidental Chlorine Release", February 1975.

REFERENCE FSAR, Section 7. 1.

Amendment No. Hl, 69 43b

1I Table 3.5.1-1 Cont'd ENGINEERED SAFEGUARDS ACTUATION SYSTEM Cont d No. of Operator action channels Min. Min. if conditions or No. of for sys- operable degree of column 3 or 4 channel s ~tem tri channels ~redundanc cannot be met

4. Reactor building spray pumps (Note 8)

a. Reactor building 30 psig instrument channel 3 (Note 6) 1 Notes 1, 5

b. Manual trip pushbutton Notes 1, 5

5. Reactor building spray valves (Note 8)

a. Reactor building 30 psig instrument channel 3 (Note 6) 1 Notes 1, 5 EMERGENCY FEEDWATER INITIATION AND CONTROL SYSTEM

1. EFW Initiation

a. Manual Note 1 45a

gE TABLE 3.5.1-1 (cont'd)

EMERGENCY FEEDWATER INITIATION AND CONTROL SYSTEM No. of Operator action channels Min. Min. if conditions or No. of for sys- operable degree of column 3 or 4 channels ~tem tni channels ~nedundanc cannot be met

b. Low Level SG A or B 4/SG 2/SG 2/SG Note 1

c. Low Pressure, SG A or B 4/SG 2/SG 2/SG Note 1, 19

d. Loss of Both MFM Pumps and PMR

> 10X Note 1

e. Loss of 4 RC Pumps Note 1, 15

f. ESAS Actuation Logic Tripped 2 Note 1

2. SG-A Main Steam Line Isolation

a. Manual Note 1

b. Low SG A Pressure Note 1, 19

3. SG-B Main Steam Line Isolation

a. Manual Note 1

b. Low SG 8 Pressure Note 1, 19 45b

TABLE 3.5. 1-1 (cont' EMERGENCY FEEDMATER INITIATION AND CONTROL SYSTEM No. of Operator action channels Min. Min. if conditions or No. of for sys- operable degree of column 3 or 4 channels ~tem tri channels ~redundanc cannot be met OTHER SAFETY RELATED SYSTEMS

1. Decay heat removal system isolation valve automatic closure and interlock system

a. Reactor coolant pressure instrument channels Notes 1, 5

b. Core flood isolation valve interlocks Notes 1, 5 45c

TABLE 3.5.1-1 (Cont'd)

OTHER SAFETY RELATED SYSTEMS No. of Operator action channels Min. Min. if conditions or No. of for sys operable degree of column 3 or 4 channels ~tern tri channels ~redundanc cannot be met

2. Pressurizer level channels 3 N/A Note 10

3. Emergency Feedwater Flow channels . 2/S.G. N/A 1 0 Note 10 4 RCS subcooling margin monitors 2 N/A 1 -0 Note 10

5. Electromatic relief valve flow monitor 2 N/A Note ll

6. Electromatic relief block valve position indicator N/A Note 12 Pressurizer code safety valve flow monitors 2/valve N/A 1/valve Note 10

8. Degraded Voltage Monitoring

a. .4. 16KV Emergency Bus Undervoltage 2/Bus 1/Bus 2/Bus Note 14

b. 460V Emergency Bus Undervoltage "1/Bus 1/Bus 1/Bus Notes 13, 14

9. =-

Chlorine Detection Systems Notes 17, 18 "Two undervoltage relays per bus are used with a coincident trip logic (2-out-of-2)

Amendment No. I, 88, 69 45d

TABLE 3.5.1-1 Cont'd)

NOTES: 1. Initiate a shutdown using normal operating instructions and place the reactor in the hot shutdown condition within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months if the requirements of Columns 3 and 4 are not met.

2. When 2 of 4 power range instrument channels are greater than 10'ated power, hot shutdown is not required.

When 1 of 2 intermediate range instrument channels is greater than 10- amps, hot shutdown is not required.

For channel testing, calibration, or maintenance, the minimum number of operable channels may be two and a degree of redundancy of one for a maximum of four hours, after which Note 1 applies.

5. If the requirements of Columns 3 or 4 cannot be met within an additional 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , place the reactor in the cold shutdown condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

6. The minimum number of operable channels may be reduced to 2, provided that the system is reduced to 1 out of 2 coincidence by tripping the remaining channel. Otherwise, Specification 3.3 shall apply.

7. These channels initiate control rod withdrawal inhibits not reactor trips at <10%%u'ated power.

Above 10K rated power, those inhibits are bypassed.

8. If any one component of a digital subsystem is inoperable, the entire digital subsystem is considered inoperable. Hence, the associated safety features are inoperable and Specification 3.3 applies.

9. Deleted

10. With the number of operable channels less than required, either restore the inoperable channel to operable status within 30 days, or be in hot shutdown within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

With the number of operable channels less than required, isolate the electromatic relief valve within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , otherwise Note 9 applies.

Amendment No. SH, 60 45e

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TABLE 3.5.1-1 (Cont'd)

12. With the number of operable channels less than required, either return the indicator to operable status withi.n 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or verify the block valve closed and power removed within an additional 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . If the block valve cannot be verified closed within the additional 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , de-energize the electromatic relief valve power supply within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

13. Channels may be bypassed for not greater than 30 seconds during reactor coolant pump starts. If the automatic bypass circuit or its alarm circuit is inoperable, the undervoltage protection shall be restored within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , otherwise, Note 14 applies.

14. With the number of channels less than required, restore the inoperable channels to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in hot shutdown within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in cold shutdown within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

15. This trip function may be bypassed at up to lOX reactor power.

16. This trip function may be bypassed at up to 20K reactor power.

17. With no channel operable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months restore the inoperable channels to operable status, or initiate and maintain operation of the control room emergency ventilation system in the recirculation mode of operation.

18. With one channel inoperable, restore the inoperable channel to operable status within 7 days or within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months initiate and maintain operation of the control room emergency ventilation system in the recirculation mode of operation.

19. This function may be bypassed below 750 psig OTSG pressure. Bypass is automatically removed when. pressure exceeds 750 psig.

Amendment No. 88, Sg, 69 45f

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Other channels are subject only to "drift" errors induced within the instrumentation itself and, consequently, can tolerate longer intervals between calibrations. Process system instrumentation errors induced by drift can be expected to remain within acceptable tolerances if recalibration is performed once every 18 months.

Substantial calibration shifts within a channel (essentially a channel failure) will be revealed during routine checking and'testing procedures.

Thus, minimum calibration frequencies for the nuclear flux (power range) channels, and once every 18 months for the process system channels is considered acceptable.

~Testin On-line testing of reactor protective channel and EFIC channels is required once every 4 weeks on a rotational or staggered basis. The rotation scheme is designed to reduce the probability of an undetected failure existing within the system and to minimize the likelihood of the same systematic test errors being introduced into each redundant channel.

The rotation schedule for the reactor protective channels is as follows:

Channels A, B, C, D Before Startup if shutdown greater than 24 hours Channel A One week after startup Channel B Two weeks after startup Channel C Three weeks after startup Channel D Four weeks after startup The reactor protective system instrumentation and EFIC test cycle is continued with one channel's instrumentation tested each week. Upon detetion of a failure that prevents trip action, all instrumentation associated with the protective channels will be tested after which the rotational test cycle is started again. If actuation of a safety channel occurs, assurance will be required that actuation was within the limiting safety system setting.

The protective channels coincidence logic and control rod drive trip breakers are trip tested every four weeks. The trip test checks all logic combination and is to be performed on a rotational basis. The logic and breakers of the four protective channels shall be trip tested prior to startup and their individual channels trip tested on a cyclic basis.

Discovery of a failure requires the testing of all channel logic and breakers, after which the trip test cycle is started again.

Amendment No. 25

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The equipment testing and system sampling frequencies specified in Table

4. 1-2 and Table 4. 1-3 are considered adequate to maintain the status of the equipment and systems to assure safe operation.

REFERENCE.

FSAR Section 7.1.2.3.4 68a

TABLE 4.1-1 (Cont'd)

Channel Descri tion Check Test Calibrate Remarks

30. Decay heat removal S(1)(2) M(1)(3) (1) Includes RCS Pressure Analog Channel system isolation valve automatic closure and (2) Includes CFT Isolation Valve Position inter lock system (3) Shall also be tested during refueling shutdown prior to repressurization pressure greater than 300 but less tlPZn 420 psig.

31. Turbine overspeed trip mechanism

32. Diesel generator protective M NA relaying starting interlocks and circuitry

33. Off-site power undervoltage M R(1) R(1) (1) Shall be tested during refueling and protective relaying shutdown to demonstrate selective interlocks and circuitry load shedding interlocks function during manual or automatic transfer of Unit 1 auxiliary load to Startup Transformer No. 2.

34. Borated water storage tank level indicator

35. Reactor trip upon loss of M PC main feedwater circuitry Amendment No. 4, ZS, SS, 88, 61 72

TABLE =4. 1-1 '(Cont')

Channel Descri tion Check Test Calibrate Remarks

36. Boric acid addition tank

a. Level channel NA NA

b. Temperature channel M NA

37. Oegraded voltage monitoring M

38. Sodium hydroxide tank level NA NA indicator

39. Incore neutron detectors M(l) NA (1) Check functioning

40. Emergency plant radiation M(l) NA (1) Battery check instruments

41. Reactor trip upon turbine M PC trip circuitry

42. Strong motion acceleographs g(1) (1) Battery check

43. ESAS manual trip functions

a. Swtiches & logic NA

b. Logic NA

44. Reactor manual trip NA

45. Reactor building sump level NA NA

46. EFM flow indication M NA Amendment No. gS, 89, Sg, 68, 61 72a

TABLE -4.1-1 (Cont'd)

Channel Descri tion Check Test Calibrate Remarks

47. RCS subcooling margin NA monitor

48. Electromatic relief valve D flow monitor

49. Electromatic relief block D NA valve position indicator

50. Pressurizer safety valve D flow monitor

51. Pressurizer water level D. NA indicator

52. Control room chlorine detector

53. EFW initiation

a. Manual NA

b. SG low level, SGA or B S

c. Low pressure SGA or B S

d. Loss of both MFW pumps S and PWR > 10K Amendment No. gS, 89, Sg, 69 72b

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TABLE 4.1-1 (Cont'd)

Channel Descri tion Check Test Calibrate Remarks

e. Loss of 4 RC pumps NA

f. ESAS automatic NA NA logic tripped

54. SGA main steam line isolation

a. Hanual

b. SGA pressure low

55. SGB main steam line isolation

a. Hanual NA

b. SGB pressure low S

56. EFM valve commands (Vector)

a. SG A pressure low

b. SG B pressure low

c. SG pressure difference SG A pressure>

SG B pressure SG B pressure>

SG A pressure 72c

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TABLE 4.1-1 (Cont'd)

Channel Descri tion Check Test Calibrate Remarks

d. SG A high range level S high-high

e. SG B high range level S high-high NOTE

- Shift - Twice every 18 months

- Quarterlyper S Each T/M Meek R Once M Meekly Q

- Prior to each PC Prior to going Critical if not M

Monthly P done within previous 31 days D

- Daily startup if not done NA - Not Applicable previous week B/N - Every 2 months 72d

P C

TABLE 4.1-2 (Continued)

Minimum E ui ment Test Fre uenc Item Test ~Fre uenc Decay heat removal Functioning Every 18 months system isolation valve automatic closure and isolation system

12. Flow limiting annulus Verity, at normal One year, two years, on main feedwater line operating conditions, three years, and every at reactor building that a gap of at least five years thereafter penetration 0.025 inches exists measured from date of between the pipe and initial test.

the annulus.

13. Main steam isolation a. Exercise through a. quarterly valves approximately lOX travel

b. Cycle b. Every 18 months

14. Main feedwater Exercise thr'ough a. quarterly isolation valves ap'proximately 10'ravel

b. Cycle b. Every 18 months

15. Re'actor internals Demonstrate oper ability Each refueling shutdown.

vent valves by:

a ~ Conducting a remote visual inspection of visually accessible surfaces of the valve body and disc sealing faces and evaluating any observed surface irregularities.

b. Verifying that the valve is not stuck'n an open position, and C. Verifying through manual actuation that the valve is fully open with a force of < 400 lbs (applied vertically upward);

Amendment No. g, Zl, gS, order 73a dated 4/20/81

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4.8 EMERGENCY FEEDWATER PUMP

~A1i bill Applies to the periodic testing of the turbine and electric motor driven emergency feedwater pumps.

~gb 'ective To verify that the emergency feedwater pump and associated valves are operable.

S ecification 4.8.1 Each EFW train shall be demonstrated operable:

a ~ By verifying on a STAGGERED TEST BASIS:

l. at least once per 31 days or upon achieving hot shutdown following a plant heatup and prior to criticality, that the turbine-driven pump starts, operates for a minimum of 5 minutes, and develops a discharge pressure of

> 1200 psig at a flow of > 500 gpm through the test loop flow path.

2. at least once per 31 days by verifying that the motor driven EFM pump starts, operates for a minimum of 5 minutes and develops a discharge pressure of > 1200 psig at a flow of > 500 gpm through the test loop flow path.

b. At least once per 31 days by verifying that each valve (manual, power operated or automatic) in each EFW flowpath that is not locked, sealed, or otherwise secured in position, is in its correct position.

C. Prior to exceeding 280'F reactor coolant temperature and after any EFM flowpath manual valve alterations by veritying that each manual valve in each EFW flowpath which, if mis-positioned may degrade EFW operation, is locked in its correct position.

d. At least once per 92 days by cycling each motor-operated valve in each flowpath through at least one complete cycle.

e. At least once per 18 months by functionally testing each EFM train and:

l. Verifying that each automatic valve in each flowpath actuates automatically to its correct position on receipt of an actuation signal.

Amendment No. ZS, 50 105

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. 2- Verifying that the automatic steam supply valves associated with the steam turbine driven EFW pump actuate to their correct positions upon receipt of an actuation signal.

3. Verifying that the motor-driven .EFW pump starts automatically upon receipt of an actuation signal.

4. Verifying that feedwater is delivered to each steam generator using the electric motor-driven EFW pump.

5. Verifying that the EFW system can be operated manually by over-riding automatic actuation signals to the EFW valves.

Bases The monthly testing frequency will be sufficient to verify that both emergency feedwater pumps are operable. Verification of correct operation will be made both from the control room instrumentation and direct visual observation of the pumps. The cycling of the emergency valves assures valve operability when called upon to function.

The functional test, performed once every 18 months, will verify that the flow path to the steam generators is open and that water reaches the steam generators from the emergency feedwater system. The test is done during shutdown to avoid thermal cycle to the emergency feedwater nozzles on the steam generator due to the lower temperature of the emergency feedwater.

The automatic actuation circuitry testing and calibration will be performed per Surveillance Specification 4. 1, and will be sufficient to assure that this circuitry will perform its intended function when called upon.

Amendment No. ZS, 50 105a

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1 ly II

II

ML17325B680

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