ML20092A369

Proposed Tech Specs Revising Accident Monitoring Instruments & post-accident Monitoring Instrumentation Tables

ML20092A369
Person / Time
Site:	Crane
Issue date:	06/13/1984
From:	GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20092A354	List: ML20092A348 ML20092A355 ML20092A369
References
TASK-2.B.3, TASK-TM NUDOCS 8406190173
Download: ML20092A369 (8)

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LIST OF TABLES TABLE TITLE PAGE

.1.2 Frequency Notation 1-8 2.3-1 Reactor Protection System Trip Setting Limits 2-9 3.5-1 Instruments Operating Conditions 3-29 3.5-2 Accidenc Monitoring Instruments 3-40c 3.5-3 Fost Accident Monitoring Instrumentation 3-40d 3.16-1 Safety Related Shock Suppressors (Snubbers) 3-65 3.18-1 Fire Detection Instruments 3-87 3.21-1 Radioactive Liquid Effluent Monitoring Instrumentation 3-97 3.21-2 Radioactive Liquid Effluent Monitoring Instrumentation 3-101 3.23-1 Radiological Environmental Monitoring Program 3-122 3.23-2 Reporting Levels for Radioactivity Concentration 3-126 in Environmental Samples 4.1-1 Instrument Surveillances Requirements 4-3 4.1-2 Minimum Equipment Test Frequency 4-8 4.1-3 Minimum Sampling Frequency 4-9 4.1-4 Post Accident Monitoring Instrumentation 4-10b l

4.2-2 Surveillance Capsules Insertion and Withdrawal Schedule 4-27a 4.19-1 Minimum Number of Steam Generators to be

't-84 Inspected During Inservice Inspection 4.19-2 Steam Generator Tube Inspection 4-85 4.21-1 Radioactive Liquid Effluent Monitoring 4-88 Instrumentation Surveillance Requirements l

4.21-2 Radioactive Gaseous Effluent Monitoring 4-91 Instrumentation Surveillance Requirements D

4.22-1

. Radioactive Liquid Waste Sampling & Analysis Program 4-98 4.22-2 Radioactive Gaseous Waste Sampling & Analysis Program 4-106 4.23-1 Maximum Valves for the Lower Limits of Dectection (LLD).

4-118 Amendment % 72 vi V

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' 3.5.5 ACCIDENT MONITORING INSTRUMENTATION t

Applicability Applies.to the operability requirements for the instruments identified l

l in Table 3.5-2 and Taale 3.5-3 during STARTUP, POWER OPERATION and HOT STANDBY.

Objectives To assure operability of key instrumentation useful in diagnosing situations which could represent or lead to inadequate core cooling or evaluate and predict the course of accidents beyond the design I

basis.

Specification 3.5.5.1 The minimum nunber of channels identified for the instruments in Table 3.5-2, shall be OPERABLE.

With the number of instrumentation channels less than the minimum required, restore the inoperable i

channel (s) to OPERABLE status within seven (7) days (48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> for pressurizer level) or be in at least HOT SHUTDOWN within the next six (6) hours and in COLD SHUTDOWN within an additional 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. Prior to starttp following a COLD SHUTDOWN, the minimum nuaber of channels shown in Table 3.5-2 shall be operable.

3.5.5.2 The channels identified for the instruments specified in Table 3.5-3 I

shall be OPERABLE. With the number of instrumentation channels less than required, restore the inoperable channel (s) to OPERABLE in accordance with the action specified in Taole 3.5-3.

Bases The Saturation Margin 2 nitor provides a quick and reliable means for determination of saturation temperature margins. Hand calculation of saturation pressure and saturation temperature margins can be easily and quickly performed as an alternate indication for the Saturation l

Margin Monitors.

i Discharge flow from the two (2) pressurizer code safety valves and

' the PORV is measured by differential pressure transmitters connected across elbow taps downstream of each valve. A delta-pressure indication from each pressure transmitter is availaole in the control room to indicate code safety or relief valve line flow. An alarm is t

. also provided in the control room to indicate that discharge from a pressurizer code safety or relief valve is occurring.

In addition, an ' acoustic monitor is provided to detect flow in the PORV discharge line. An alarm is provided in the control rocm for the acoustic monitor.

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The Emergency Feedwater System is provided with two channels of flow instrumentation on each of the two discharge lines. Local flow indication is also available for the emergency feedwater system.

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Although the pressurizer has multiple level indications, the separate indications are selectable via a switch for display on a single display.

Pressurizer level, however, can also be determined via the patch panel and the computer log.

In addition, a second channel of pressurizer level indication is available independent of the NNI.

Although the instruments identified in Table 3.5-2 are significant in diagnosing situations which could lead to inadequate core cooling, loss of any one of the instruments in Table 3.5-2 would not prevent continued, safe, reactor operation. Therefore, operation is justified for up to 7 days (48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> for pressurizer level). Alternate indications are availaole for Saturation Margin Monitors using hand calculations, the PORV/ Safety Valve position monitors using discharge line thermocouple and Reactor Coolant Drain Tank indications, and for EFW flow using Steam Generator level and EFW pump discharge pressure. Pressurizer level has two channels, one channel from NNI (3 O/P instrument strings through a single indicator) and one channel independent of the NNI. Operation with the above pressurizer level channels out of service is permitted for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. Alternate indication would be available through the plant computer.

Manitors for containment pressure, containment water level, containment hydrogen level and various high range radiation monitors are useful to evaluate and predict tne course of accidents which go beyond tne plant design basis (See Table 3.5-3).

These instruments should ce maintained for that purpose.

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.iLBLE'3.5-3 Post Accident Monitoring Instrumentation FUNCTION IN S T R UMEN T S REQUIRED NUMBER MINIMUM NUMBER ACTION OF Cil ANNEL S OF CHANNELS 1.

High Range Noble Gas E f fluent a.

Condensei Vacuum Pump Exhaust 1

1 A

(RH-A5-Hi) b.

Condenser vacuum Pump Exhaust 1

1 A

( R M-G2 5 )

c.

Auxiliary and fuel llandling 1

1 A

Guilding E xhaus t ( RH-AB-Hi) d.

Reactor Building Purge Exhaust 1

1 A

( R M-A9-H i) e.

Reactor Building Purge Exhaust 1

1 A

ug o

(RH-G24) f.

Main Steam Lines Radiation 1

1 A

(RM-G26/RH-G27) each OTSG each UTSG 2.

Containment liigh Range Radiation 2

2 A

(RH-G22/23) 3.

Containment Pressure 2

1 0

4.

Containment Water Level 2

1 0

2 1

0 5.

Containment Hydrogen

p TABLE 3.5-3 (Continued)

ACTIONS A.

With the number of OPERABLE Channels less than required by the Minimum channels OPERABLE requirements:

1.

either restore the inoperable Channel (s) to OPERABLE status within 7 days of the event, or 2.

prepare and submit a Special Report within 30 days following the event outlining the action taken, tne cause of the inoperability and the plans and schedule for restoring tne system to OPERABLE status.

B.

1.

With the number of OPERABLE accident monitoring instrumentation channels less than the required CHANNELS OPERABLE requirements restore the inoperable channel (s) to OPERABLE status within 30 days or. be in at least HOT SHUTDOWN within the next 12 holes.

2.

With the number of OPERABLE accident monitoring instrumentation channels less than the MINIMUM CHANNELS OPERABLE requirements, restore tne inoperable channel (s) to OPERABLE status witnin 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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SURVEILLANCE STANDARDS Specified intervals may be adjusted plus or minus 25 percent to accommodate normal test schedules.

4.1 OPERATIONAL SAFETY REVIEW Applicability Applies to items directly related to safety limits and limiting conditions for operation.

Objective To specify the minimum frequency and type of surveillance to be applied to unit equipment and conditions.

Specification 4.'l.1 The minimum frequency and type of surveillance required for reactor protection system and engineered safety feature protection system instrumentation when the reactor is critical shall be as stated in Table 4.1-1.

l 4.1.2 Equipment and sampling test shall be performed as detailed in Tables 4.1-2 and 4.1-2.

4.1.3 Each post accident monitoring intrumentation channel shall be demonstrated OPERABLE by the performance of the check, test and calibration at the frequencies shown in Table 4.1-4.

Bases Check Failures such as blown instrument fuses, defective indicators, or faulted amplifiers which result in " upscale" or "downscale" indication can De easily recognized by simple observation of die functioning of an Instrument or system. Furthermore, such failures are, in many cases, revealed by alarm or annuciator action. Comparison of output and/or state of independent channels measuring the same variable supplements this type of built-in surveillance.

Based on experience in operation of both conventional and nuclear systems, when the unit is in operation, the minimum checking frequency stated is deemed adequate for reactor system instrumentation.

Calibration Calibration shall be performed to assure tne presentation and acquisition of accurate information. The nuclear flux (power range) channels amplifiers shall be checked and calibrated if necessary, every shift against a heat balance standard. The frequency of heat calance checks will assure that the difference between the out-of-core instrumentation and tne heat balance remains less than 4%.

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T ABLE 4.1-1 (Centinuad)

Cil ANN E L DE SCill PI I DN CllECK TEST CALIDRATE ItC H Alt K S 20.

Radiation Honitoring Systems

• W(1)(3)

H(3)

Q(2)

(1)

Using the installed check source w ho sil* '

background is less than twice the e x pe c t c al increase in cpm which would result from the check source alone.

Gackground reodings greater than this value are sufficient in themselves to show that the monitor is functioning.

(2)

Except arco gamma radiation monitors It H-C 6, RH-G7, and itH-GU, which are located in high radiation arous of the lle n c t o r Guilding.

These monituru will be calibrated quarterly or at the next scheduled reactor shutdown following thu quarter in which calibra tion would 4.

di normally be due, if a shutdown during the W

quarter does not occur.

(3)

Surveillances are required to be l

performed only when containment integrity is required.

This applies to monitors which initiate containment isolation only.

29.

liigh,and Low P ressure NA NA R

Injection Systems:

Flow Channels.

4 Duas not include the monitors covered under specification 3.5.5.2 and 4.1.3

_ _ _ _ _ = _ = _ _ _ _ _ _ _ _ _ _.

T ABL E 4.1-4 POST ACCIDENT HONITORING IN S T RUMEN T A T I ON FUNCTION I N S T R UMEN T S CHECK TEST CALIBRATE HEMAHXS 1.

Noble Gas Errluent a.

Condenser Vacuum Pump Exhaust (RH-AS-Hi) W H

R (1)

Using the installed check source when background is less than twice the expected increase in cpm which would result.from the check source alone.

Dackground readings greater than this value are sufficient in themselves to show that this monitor is functioning.

b.

Condenser Vacuum Pump E xhaun t (RM-G2 5)

W(1)

H R

c.

Aux. & f uel Handling Building Exhaust W

H R

pg (RH-AD-Hi)

DF d.

Reactor Building Purge Exhaust W

H R

(RM-A9-HI) e.

Reactor Building Purge Exhaust W( 1)

H R

(RH-G24) r.

Main Steam Lines Radiation (RM-G26/

W(1)

H R

RM-C27) 2.

Containment High Range Radiation W

H R

(RH-C 22/23) 3.

Containment Pressure W

N/A R

4 Containment Water Level W

N/A R

5.

Containment Hydrogen W

H R

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