ML20092L994

Proposed Tech Spec Change 84-05 Re High Water Level Scram Discharge Tanks & Reactor Protection Sys Instrumentation Functional Tests

ML20092L994
Person / Time
Site:	Pilgrim
Issue date:	06/26/1984
From:	BOSTON EDISON CO.
To:
Shared Package
ML20092L982	List: BECO-84-091, Application for Amend to License DPR-35,consisting of Proposed Tech Spec Change 84-05 Re High Water Level Scram Discharge Tanks & Reactor Protection Sys Instrumentation Functional Tests ML20092L994
References
NUDOCS 8407020097
Download: ML20092L994 (15)

Text

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ATTACHMENT A

^

Proposed Technical Specification Change Proposed Change:

Reference is made to Pilgrim Nuclear Power Station, Technical Specifications, Appendix A,. Tables 4.1.1, 4.1.2, 3.2.C, and associated Notes and Bases.

The following changes are requested to be made:

1) Table 4.1.1 change:

"High Water Level Scram Discharge Tank" to:

"High Water Level Scram Discharge Tanks" change: Group "A"Lon that same line to:

Group "D" 2)' Notes'for Table 4.1.1 change Note'2 from:

"A description of the three groups... of th'is Specification" to:

"A description of the four groups... of this Specification"

3) Table 4.1.2 change:

"High Water Level in Scram Discharge Tank" to:

"High Water Level in Scram Discharge Tanks" change: Group "A" on that same line to: Grcup "D" change: Note (6) reference in both columns to:

Note (7) 4)~ Notes for Table 4.1.2 change Note 1 from:

"A description of three groups... of this Specification" to:

"A description of four groups... of this Specification" delete from Note 6 the word "the" between "during" and " refueling outages",

add'the following note:

"7. Calibration of these devices will be performed during refueling outages."

8407020097 840626 PDR ADOCK 05000293

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5)~LBases

Pages 34, 35, 36,:37 and 38

' Replace existing pages with marked up pages provided.

6) : Table 3.2.C' change:

"1.(9)"

to:

"1 (per tank) (9)"

Reason for Change:

.- Per-NRC Confirmatory Order dated June 24, 1983, Boston Edison Company has

- permanently modified the Scram Discharge Volume (SDV) system to the criteria endorsed by'the BWR Owners' Subgroup and endorsed by the NRC Generic Safety Evaluation Report on the BHR Scram Discharge System dated December 1, 1980.

- The SDV modifications include two (2) new scram discharge instrument volumes with redundant and diverse instrumentation and replacement of the 2 inch

- diameter SDV header drain lines with 6 inch diameter drain. lines to improve

- hydraulic coupling.

The original scram discharge instrument volume tank and associated instruments have been removed from service.

1

- The SDV modification retains the two existing Control Rod Drive (CRD) Scram Discharge Headers.

However, with the SDV modification, each header will drain to a separate Scram Discharge Instrument, Volume (SDIV) (SDIV - East or SDIV -

West).

Each header has been provided with two redundant vent valves. and each instrument volume has two redundant drain valves.

Each SDIV har tuo level transmitters which provide continuous monitoring of

-level in the instrument volume over a range of approximately four (4) gallons to forty-five (45)' gallons. The level transmitters and asseclated bistable devices provide signals for reactor scram on high water in the SDIVs at approximately thirty-nine (39) gallons and for the "not-drained" alarms, at approximately four.and one half (4.5) gallons.

In addition each SDIV is provided with three resistance temperature device (RTD), heat actuated, level sensors.

One RTD in each~ tank provides a signal to the reactor manual control system for control rod withdrawal block if water level in the SDIV should reach approximately 18 gallons.

The two upper RTDs in each SDIV provide reactor scram signals on high water level (;:39 gallons) in addition to the two level transmitters.

~

Background

On July 7, 1980 the NRC forwarded model Tech. Specs. for SDV from which BECo proposed limiting conditions of operation and surveillance revisions to the PNPS Tech. Specs.

These were approved by your office and issued as Tech.

Spec. Amendment #65 dated November 10, 1982. On June 24, 1983, a Confirmatory Order was issued to BECo, confirming car proposed schedule and design modifications to the SDV system. Accompanying the confirmatory order was a set of model Tech. Specs, which we reviewed and found to be consistent with the previously issued Amendment 65 Technical Specification.

The changes being proposed at this time are administrative in nature, in that part of the

p modification to the 3DIVs consists of replacing the level switches with a new group of analog transnitter trip devices.

This in turn necessitates a nomenclature change in the Technical Specification Tables.

No other changes are considered necessary, as the other part of the modification le.. providing diverse and redundant instrumentation and installing dual SDIVs, do not involve.LCO,:survelliance, or instrumentation setpoint changes.

The incorporation of the new group of trip devices as well as an update to the bases.that describe them, constitutes the entire T.S. change proposal.

Safety Considerations:

1This change does not present an unreviewed safety question as defined in 10 CFR 50.59.

It-has been reviewed and approved by the Operations Review

~ Committee and reviewed by the Nuclear Safety Review and Audit Committee.

Significant Hazards Considerations:

It has been determined that this. amendment request involves no significant hazards consideration. Under the NRC's regulations in 10 CFR 50.92, this

. means that operation of the Pilgrim Nuclear Power Station in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

The NRC has provided guidance concerning the application of standards for

' determining whether license amendments involve significant hazards considerations by providing certain examples (48 FR 14870). One example of an amendment that is considered not likely to involve a significant hazards consideration is:

"(11)

A change that constitutes an additional limitation, restriction, or control not presently included in the Technical Specifications:

for example, a more stringent surveillance requirement."

The present Technical Specifications were written to apply to a single SDIV and its associated instrumentation.

~

With the addition of a second SDIV, the existing limitations and testing frequencies will now apply to the dual SDIV system which could be considered to be a doubling of surveillance testing.

Therefore, this change constitutes

- "a more stringent surveillance requirement" and example (11) would apply.

Schedule of Change:

This change, based upon NRC approval, will become effective when operation with the SDIV system is required.

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Attachment 81 f: ~

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Proposed-Technical'Specificatisn Change-s

'Q List of'pages affected by;this.. change:

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Pages 30, 31, 32,-33, 34, 35, 36, 37,.38,'39,and 54.

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s TABLE 4.1.1 -

iREACTOR PROTECTION SYSTEM (SCRAM)" INSTRUMENTATION FUNCTIONAL TESTS.

MINIMUM FUNCTIONAL TEST FREQUENCIES.FOR SAFETY INSTR. AND CONTROL CIRCUITS.

~

Group,.(2).

Functional Test

- Minimum Frequency (3).

Mode Switch in Shutdown ~

.A

~ Place Mode Switch in Shutdown' Each Refueling Outage-

RPS Channel Test Switch-(5)-

A Trip; Channel and Alarm

. Every 3 Months Manual. Scram

A.

Trip Channel and Alarm -

Each Refueling Outage IRM --

-High Flux C

Trip Channel ~and Alarm (4)

Once Per Week During Refueling and Before Each'Startup.

Inoperative C

Trip Channel and'Alars' Once Per Week During' Refueling and Before Each Startup',

APRM High Flux' B

Trip output' Relays (4)

. Once/ Week (7).

-Inoperative B-Trip Output Relays (4).

.Once/ Week Downscale B

Trip Output Relays.(4)

. Once/ Week Flow Biase-B Calibrate Flow Bias Signal

~ Once/ Month (1)

High Flux-(15%)

B.

Trip output Relays (4)

, 0nce Per Week During Refueling and Before Each Startup High Reactor Pressure A

Trip Channel and Alarm (1)

].

, Reactor' Low Water Level (6)

A Trip Channel and Alarm

-(1) k High Drywell Pressure A'

Trip Channel and Alarm (1)

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High Water Level in Scr.m Discharge Tank.

-D Trip Channel and Alarm Every 3 Months Turbine Condenser Low Vacuum A

Trip Channel and Alarm (1) l' Main St'eam Line High Radiation B

Trip Channel and Alarm (4)

Once/ Week Main Steam Line Isolation Valve Closure A

Trip Channel and Alarm (1)

~

j Turbine, Control Valve Fast' Closure A

Trip Channel and Alarm (1) q-ATurbine First Stage Pressure Permissive A

Trip Channel and Alarm Every 3 Months 6

s Turbire-Stop Valve-Closure A-Trip Channel and Alarm

-(1) l e

~ Reactor-Pressure: Permissive A

Trip Channel and Alarm Every 3 Months j.

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' NOTES FOR TABLE 4'.1.1 1.

Initially 5 nce Per month until exposure (M as defined on Figure 4.1.1) is 2.0 x 10 ; thereafter, according to Figure 4.1.1 with an interval not less than one month ~ nor more than three months.

The compilation of instrument failure rate data may include data obtained from other boiling water reactors for which the same design instrument operates in an environment similar to that of PNPS.

2.

A description.of the - four groups is included in the Bases of this 'Speci-fication.

3.

Functional tests are not required when the systems are not required to be operable or are tripped.

If tests are missed, they shall be performed prior to returning the systems to an operable status.

4.

.This instrumentation is exempted from the instrument channel test defini-tion..This instrument channel functional test will consist of injecting a simulated electrical signal into the measurement channels.

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Test RPS channel after maintenance.

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The water level in the reactor vessel will be perturbed and the corres-ponding level indicator changes will be monitored.

This perturbation test will' be performed every month after completion of the monthly func-tional test program.

7.

This APRM testing will be performed once per week when in the run mode.

If the reactor is out of the run mode for more than one week, the testing will'be performed as soon as practicable after returning to the run mode.

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TABLE 4.1.2 REACTOR PROTECTION SYSTEM (SCRAM INSTRUMENT CALIBRATION -

MINIMUM CALIBRATION FREQUENCIES FOR REACTOR PROTECTION INSTRUMENT CHANNELS Instrument Channel Group (1)

Calibration Test (5)

Minimum Frequency (2)

IRM High Flux C

' Comparison to APRM on Controlled Note (4)

Shutdowns

. A"P.M High Flux Output Signal B

Heat Balance

- Once every 3 Days B

Internal Power and Flow Test Each Refueling Outage Flow Bias Signal LPRM' Signal B

TIP System Traverse Every 1000 Effective Full Power Hours High Reactor Pressure A

Standard Pressure Source Every 3 Months High Drywell Press'4re A

Standard Pressure Source Every 3 Months Reactor Low Water Level A

Pressure Standard Every 3 Months High Water Level in Scram Discharge Tank D

Note (7)

Note (7)

Turbine Condenser Low Vacuum ~

A-Standard Vacuum Source Every 3 Months Main Steam Line Isolation Valve Closure A

Note (6)

Note (6)

Main Steam Line High Radiation B

Standard Current Source (3)

Every 3 Months Turbine First, Stage Pressure Pennissive A

Standard Pressure Source Every 6 Montns A

Standard Pressure Source Every 3 Months Turbine Control V:1ve Fast Closure Turbine Stop Valve Closure A

Note-(6)

Note (6)

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Reactor Pressure Permissive A

Standard Pressure Source Every 6 Montbs s

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groups is included in the bases of this Specifi-4.

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A description of four 4

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Calibration ' tests are not required when the systems are not required to be operaMe or are tripped.

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Tae current source provides an instrument channel alignment.

Calibration C

using a radiatica source shall be made each refueling outage.

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Maximum frequency required is.once per week, h

5.

Response time is not a part of the routine instrument channel test, but will be checked once per operating cycle.

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BASES:

3.1 The reactor protection system 4.1 A.

The minimum functional test-automatically initiates a reac-ing frequency used in this tor scram to:

specification is based on a reliability analysis using 1.

Preserve the integrity of the concepts developed the fuel cladding.

in reference (6). This con-cept was specifically adapt-2.

Preserve the integrity of

.ed to the one out of two the reactor coolant system.

taken twice logic of the reactor protection system.

3.

Minimize the energy which The analysis shows that the must be absorbed following sensors are primarily respon-a loss of coolant accident sible for the reliability and prevents criticality.

of the reactor protection system. This analysis makes This specification provides the use of " unsafe failure" rate limiting conditions for opera-experience at conventional

tion necessary to preserve the and nuclear power plants in ability of the system to tole-a reliability model for the rate single failures and still system. An " unsafe failure" perform its intended. function is defined as one which ne-even during periods when in-gates channel operability strument channels may be out and which, due to its na-of service because of mainte-ture, is revealed only when nance. When necessary, one the channel is functionally channel may be made inoperable tested ~ or attempts to re-for;brief intervals to conduct spond to a real signal.

required functional tests and Failures such as blown calibrations.

fuses, ruptured bourdon tubes, faulted amplifiers, Tne reactor protection system faulted cables, etc. which is of the dual channel type.

result in " upscale" or Ref. Section 7.2 FSAR. The "downscale" readings on the system is made'up of two in-reactor instrumentation are dependent trip systems, each

" safe" and will be easily having two subchannels of recognized by the operators tripping devices. Each sub-during operation because channel has an input from at they are revealed by an least one instrument channel alarm or a scram.

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which monitors a critical parameter.

The channels listed in Tables 4.1.1 and 4.1.2 are The outputs of the subchannels divided into four groups are combined in a 1 out of 2 for functional testing.

logic; i.e.,_an input signal These are:

on either one or both of the subchannels will cause a trip A.

On-Off sensors that system trip. _The outputs of provide a scram trip the trip systems are arranged function.

'so that a trip-on both systems is required B.

Analog devices coupled with bi-stable trips that provide a scram function.

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3.1 BASES'(Cont'd) 4.1 BASES (Cont'd) to provide a-reactor scram. This C.

Devices which only serve a

--system meets the intent of useful function during some

. EEI - 279 for Nuclear Power Plant restricted mode of operation,

. Protection Systems. The system such as startup or shutdown' has.a reliability greater than or for which the only prac-

'that of a 2 out of 3 system and tical test is one that can somewhat less than that of a 1 be performed at shutdown.

of 2 system.

D.

Diverse Analog Transmitter /

With the exception of the trip unit devices that provide Average Power Range Monitor alarms, trips or scram functions.

(APRM) channels, the. Inter-mediate Range Monitor (IRM)

The sensors that make up group channels of the Main Steam (A) are specifically selected Isolation Valve closure from among the whole family of

and-the Turbine Stop Valve industrial on-off sensors that closure, each subchannel has have earned an excellent reputa-one instrument channel. When tation for reliable operation.

Lthe minimum condition for During design, a goal of 0.99999 operation on'the. number of probability of success (at the operable. instrument channels 50% confidence level) was adop-

-per untripped protection trip ted'to assure that a balanced and system is met or it.if cannot adequate design is achieved. The be met and the effected pro-probability of success is primar-tection trip system is placed ily a function of the sensor fai-in a~ tripped condition, the lure rate and the test interval.

effectiveness of the protection A three-month test interval wa,s

. system is preserved; i.e., the planned for gr'oup (A) sensors.

system can tolerate a single.

This is in keeping with good failure and still perform its operating practices, and satis-intended function of scramming fics'the design goal for the

'the-reactor. Three'APRM in-logic configuration utilized in y

strument. channels.are provided

.the Reactor Protection System.

for each protection trip system.

To satisfy the long-term objec-APRM's #1 and #3 operate con-tive of maintaining-an adequate

' tacts in one subchannel and level of safety throughout the P ant-lifetime, a minimum goal of l

APRM's #2 and #3 operate con-

. tacts in the other subchannel.

0.9999 at the 95% confidence

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APRM's #4, #5, and #6 are level is proposed. With the e

arranged similarly in the (1 out of 2) X.(2) logic, this

-other protection trip system.

requires that each sensor have an availability of 0.993 at the Each protection trip system has one more APRM than'is 95% confidence Icvel. This level necessary to meet the minimum of availability may be maintained

. number required per channel. This by adjusting the test interval as l

a function of the observed fail-allows _the bypassing of one APRM ure history (6).

per protection trip system for

- maintenance, testing or calibra-(6) Reliability of Engineered stion. Additional IRM channels Safety Features as a Func-tion of Testing Frequency, have also I.M. Jacobs " Nuclear Safety,"

Vol. 9, No. 4, July-Aug.

1968, pp. 310-312 35 l

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3.1 BASES (Cont'd) 4.I BASES (Cent'd) been provided to allow for by-To facilitate the implementation passing of one such channel, of this technique, Figure 4.1.1 The bases for the scram setting is provided to indicate an appro-for the IRM, APRM, high reactor priate trend in test interval.

pressure, reactor low water level, The procedure is as follows:

.MSIV, closure, generator load rejection, turbine stop valve 1.

Like sensors are pooled closure and loss of condenser into one group for the vacuum are discussed in Speci-purpose of data acquisition.

fication 2.1 and 2.2 2.

The factor M is the exposure hours and is equal to the

-Instrumentation (pressure number of sensors in a switches) for the drywell are provided to detect a loss of gy up, n, times the clapsed time T (M = nT).

coolant accident and initiate.

the core standby cooling equip-3.

The accumulated number of ment. A high drywell pressure unsafe failutes is plotted scram is provided at the same as an ordinate against M setting as the core cooling as an asseissa on Figure systems (CSCS)' initiation to 4.1.1 minimize the energy which must be accomodated during a loss 4.

.After a trend is established, of coolant accident and to the appropriate monthly test prevent return to criticality.

interval to satisfy the goal This instrumentation is a will be the test interval to backup to the reactor vessel the left of the plotted r

water level instrumentation.

points.

High radiation levels in the 5.

A test interval of one month main steam line tunnel-above will be used initially un-that due to the normal nitro-til a trend is established.

gen and oxygen radioactivity

.is an indication of leaking Group (B)' devices utilize an ana-fuel. A scram is initiated log sensor followed by an ampli-whenever such radiation level fier and a bi-stable trip circuit.

exceeds seven times normal The sensor and. amplifier are ac-background. The purpose of tive components and a failure is this scram is to reduce-the almost always accompanied by an source of such radiation to alarm and an indication of the extent necessary to prevent the source of trouble.

In the excessive turbine contamina-event of failure, repair or sub--

tion. Discharge of exce^ssive stitution can start immediately.

amounts of radioactivity to-

'An "as-is" failure is one that the site environs is prevented

" sticks" mid-scale and is notc by the~ air ejector off gas capable of going either up or down

. monitors which cause an isola-in response to an out-of-limits

-tion of the main condenser, input. This type of failure for off gas line'.

analog devices is a rare occur-rence and is detectable by an A reactor mode switch is pro-Perator who observes that one vided which actuates or by-

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passes the various scram func-sicnal does not track the other-tions appropriate to the par-three. For purpose of analysis, ticular plant operating status.

Ref. Section 7.2.3.7 FSAR.

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4.1' BASES (Cont'd) i T

The IRM systemiand APRm (15%).

it-is assumed.that this rare failure will be detected within scram provide. protection atainst exc.essive_ power levels-two hours.

tand short reactor periods 7 n the i

startup and' intermediate power The bi-stable trip circuit-which is a part of the Group (B) de-

~ ranges.

vices can sustain unsafe failures The control' rod drive' scram sys-which are revealed.only on test.

3 tem is' designed so that all of_

.A study was conducted of the in-Lthe; water which is discharged strumentation channels included M

from the reactor-by-a scram.

in the Group (B) devices to

.,can beJaccomodated in the dis-

. calculate their " unsafe" failure

charge piping.

rates. The anaiog devices (sen-sors and amplifiers) are predict-The tiwo scram discharge volumes accommodate _in excess of,39 gal-ed to have an unsafe _ failure. rate

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of less than 20 X 10 failure /

lons of: water each and'are at the_

hour. -The bi-stable trip cir-

' low points of the' scram discharge cuits.are. predicted to have an piping. No_ credit was taken for ithese:volumsa in-the: design of the unsafe,gailurerateoflessthan 2 X 10 failures / hour. Con-

-discharge piping as concerns the

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sidering the two hour monitoring

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= amount.ofLwater-which must be_

interval for the analog devices accommodated during a scram, as= assumed above, and a weekly

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test intervalTfor the bi-stable

During normal operation'the scram trip circuits, the design reli-discharge volume system is empty; ability goal of 0.99999 is at-

' however, should it. fill: with water, tained with ample margin.

,the water discharged-to the-piping

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icould,not be accommodated which

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The bi-stisble devices are moni-

-wo'uld result in slow' scram times.

tored during plant operation to.

'or partial control rod insertion, record.their failure history and To preclude this occurrence..redun_

establish a test interval using dant and-diverse level detection the curve of. Figure 4.1.1.

There devices in the_ scram discharge _

are numerous identical bi-stable

-instrument. volumes'have been pro-devices used throughout the

.vided which will: alarm when water plant's instrumentation system.-

. level reaches'4.5 gallons, ini_.

Therefore,.significant data'on tiate'a'controlfrod block at 18

-the failure rates for the bi-1 gallons, and scram the reactor

stable devices should be accumu-twhen the water level reaches 39

. gallons. As-indicated above, lated rapidly.

there~is~ sufficient-volume-in The frequency of calibration of

-the piping to accommodate the the APRM Flow Biasing Network

scram without-impairment of has been established as each t-

-the~ scram-times or amount of-

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. insertion-of the' control rods.

This function shuts-the reac-4

' tor down while_ sufficient-volume remains to! accommodate-the. discharged water and. pre-cludes the situation in which

-a scram would be requested but

-not be able 37 m

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3.1 BASES (Cont'd).

4.1 BASES (Cont'd) to perform its function ade-refueling outage. The flow bias-quately.

ing network is functionally test-ed at least once per month and, a source range monitor (SRM) in addition, cross calibration sytem is also provided to sup-checks of the flow input to the 3-ply additional neutron levet flow biasing network can be information.during start-up made during the functional test but has no scram functions.

by direct meter reading. There Ref. Section 7.5.4 FSAR.

are several instruments which The APRM's cover the " Refuel"

.must be calibrated and it will and "Startup/ Hot Standby" take several days to perform the

. modes. THus, the.IRM and calibration of the entire network.

APRM 15% scram are required While the calibration is being in the " Refuel" and "Startup/

performed, a zero flow signal Ect Stsadby" mode's.

In the will be sent to half of the pcwer range the APRM system APRM's resulting in a half scram provides the required pro-and rod block condition. Thus, tection. Ref..Section 7.5.7 if the calibration were perform-FSAR. Thus, the IRM system ed during operation, flux shap-

.is not required in the "Run" ing would not be possible. Based mode.

on experience of other generating stations,' drift of instruments, The high pressure high drywell such as those in the Flow Bias-

. pressure, reactor low water ing Network, is not significant y.

level and scram discharge and therefore, to avoid spurious volume high level scrams are scrams, a calibration frequency required for Startup/ Hot Stand-of each refueling outage is by and Run modes' of plant established.

operation.

They are, there-fore, required to be opera-Group (C) devices are active only tional for these modes of during a given portion of the reactor operation.

operational cycle. For example, the IRM is active during startup The-requirements to have the and inactive during full power scram functions as indicated operation. Thus, the only test in Table'3.1.1-operable in that is meaningful is the one the Refuel mode is to assure performed just prior to shut-down that shifting to the Refuel' or startup; i.e.,

the tests that mode during reactor power are performed just prior-to use operation does not diminish of the instrument.

the need for the reactor Group (D) devices, while similar Protection system.

in description to those in Group

~(B), are different in use because The turbine condenser low they (the analog transmitter / trip vacuum scram is only_ reouired unit devices) provide alarms, during power operation.and trips or scram functions.

must be bypassed to start up An availability analysis is detailed the unit. Below 305 psig in NED0-21617 (4/77).

turbine first stage pressure

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(45% of rated), the scram surveillance frequencies for the SDV system instrumentation is detailed in Amendment Number 65.'

NRC con-currence with this surveillance pro-38 L

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3.1 BASES (Cont'd) 4.1 BASES (Cont'd) signal due to turbine'stop gram is contained in the Safety valve closure is bypassed Evaluation Report and its associated because flux and pressure Technical Evaluation Report (TER-C-scram are adequate to pro-5506-66) dated 11/10/82*

tect the reactor.

Calibration frequency of the The requirement that the instrument channel is divided

_IRM's be inserted in the core into two groups. These are as when the APRM's read 2.5 in-follows:

dicated on the scale assures that there is proper. overlap 1.

Passive type indicating in the neutron monitorin8 devices that can be compared systems and thus that ade-with like units on a contin-quate coverage is provided nous basis.

for all ranges of reactor operation.

2.

Vacuum tube or semiconductor devices and detectors that The provision of an APRM drift or lose sensitivity.

scram at 15% design power in the " Refuel" and " Start-Experience with passive type in-up/ Hot Standby" modes and struments in generating stations the backup IRM scram at and substations indicates that 120/125 of full scal as-specified calibrations are ade-sures that there-is proper quate. For those devices which overlap in the neutron employ amplifiers, etc., drift monitoring systems, and, specifications call for drift thus, that adequate cover-to be less than.04% month; age-is provided for all i.e., in the period of a month ranges of reactor opera-a drift of 4% would occur and tion.

thus providing for adequate mar-gin. For the APRM system drift of electronic apparatus is not the only consideration in deter-mining a calibration frequency.

Change in power distribution and loss of chamber sensitivity dic-tate a calibration every seven days.

Calibration on this fre-quency assures plant operation at or below thermal limits.

A comparison of Tables 4.1.1 and 4.1.2 indicates that two instru-ment channels have not been in-cluded in the latter Table.

These are: mode switch in shut-down and manual scram. All of the devices or sensors associated with these scram functions are simple on-off switches and, hence, calibration during opera-tion is not applicable, i.e.,

the switch is either on or off.

39

h

- k PNPS R.:

TABLE 3.2.C M

INSTRUMENTATpNTilAT'INITIATESR0D-BLOCKS E

..f

-Minimum Il of P'

Operalile Instrument'

~

Channels Per Tri L y gems (1)

Instrument.

Trip Level Setting S

(0.58W+50%)(MELPD '

(2)

_FRP:

2 APRM tipscale (Flow Biased)

(

2 APRM Downscale

.2.5 indicated on scale 1 (7)'

Rod Block Monitor:

(0.65-V +'42%)' FRP ' (2)

(Flow Biased)

,HFI.PD,

I (7).

Rod Block Monitor-5/125 ot in11 scale i

Downscale.

3 IRM Downscale.(3) 5/125 of full scale 3

IRM Detector not in (8) 4 Startup Position 3

'IRM Upscale

$108/125 of full scale 2 (5)

SRM Detector not in (4) t Startup Position 5

2 (5) (6)

SRM Upscale 5 10 counts /sec.

i 1 (per tank) (9)

Scram Discharge Volume 518 gallons.

Water I.evel-liigh

. - -. ~.

-