ML20118A389

Proposed Tech Specs 3/4.1 & 3/4.2 Re Instrumentation Requirements

ML20118A389
Person / Time
Site:	Duane Arnold
Issue date:	08/25/1992
From:	IES UTILITIES INC., (FORMERLY IOWA ELECTRIC LIGHT
To:
Shared Package
ML20118A388	List: ML20118A389 NG-92-3686, Forwards Request for TS Change RTS-186A to License DPR-49, Revising TS Sections 3/4.1 & 3/4.2 (Instrumentation Requirements),Per NRC & BWROG Review of 911219 Request for TS Change 186
References
NUDOCS 9209020334
Download: ML20118A389 (3)

Text

DAEC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RE('UIREMENT i

3.1 REACTOR PROTECTION SYSTEM i 4.1 REACTOR PROTECTION SYSTEM

,1NSTRUMENTATION

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INSTRUMENTATION l

A.

As a minimum, the reactor A1 Each reactor protection 9yst%n i

protection system instrumentation inntrurantati on channel shall be l

channels shown in Table 3.1-1 demonstrated OPERABLE by the shall be OPERABLE with the perf orstance of the CHANNEL CHECK, PROTECTIVE INSTRUMENTATION CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for RESPONSE TIME an shown in Tablo 3.1-2.

the OPEP.t. TING MODES and at the frequencies shown in Table 4.1-1.

'Ihe designed system response timen from the opening of the l

2.

Response time measu*emouts (from sensor contact up to and actuation of set.sor contacts or including the opening of the trip trip point to de-energization of actuatnr contacts shall not scram soltnold relay) ers not part exceed 50 milliseconds.

of the normal instrument

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calibration.

The reactor trip A

system response time of eta.h y licability reactor trip functiot, shall be As shown in Table 3.1-1, demonstrated to be within its limit once por operating cycle.

Action:

Each test shall include at leact one logic train such that both 1.

With one channel required by logic traine are tested at least Table 3.1-1 inoperable in one or once per 36 months and one channel more Trip Functions, place the per function such that all inoperable channel (s) and/or that cParnels are tasted at least once trip system in the tripped every N ;imes 18 v7nths where H is l

condition

• within 12 houce.

the t-tal number of redundant channels in a specific reactor 2.

With two or more channels trip function.

required by Table 3.1-1 inoperable in one or more Trip runctions:

a.

Within one hour, verify sufficient channels remain OPERABLE or tripped

• to maintain trip capability in the Trip Functions, and b.

Within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, placc the inoperable channel (s) in one trip system and/or that trip system **

in the tripped condition *, and c.

Within 12 hoars, restore the inoperable channels in the other trip system to an OPERABLE status or tripped *,

Otherwise, take the ACTION required by Table 3.1-1 for the Trip Function.

+ An inoperable channel or trip system need not be placed in the tripped condition where this would cause the Trip Function to occur.

In these cases, if the inoperable channel is not restored to OPERABLE status within the required time, the ACTION required by Table 3.1-1 for that Trip Function shall be taken.

• • This ACTION applies to that trip system with the most inoperable channels; if both systems have the same number of inoperable channels, the ACTION can be applied to either trip. system.

RTS-186A 9209020334 920825 3.1-1 08/92 PDR ADOCK 0$000331 P

PDR

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reactor pressure, reactor low water level, MSIV closurc, generator load rejection and lurbine stop valve closure are diecussed in Specifications 2.1 and 2.2.

Instrumentation (pressure switches) for the devwell are provided to detect a loss-of-coolant accident and initiate the emergency core cooling equipment.

A high daywell pressure-scram is provided at the same setting as the emergency core cooling systems (ECCS) initiation to minimize the energy which must be accommodated during a loss-of-coolant accident and to prevent return to criticality.

This instrumentation is a backup to the reactor vessel water level instrumentation.

The reactor water level trip settings are referenced to the " top of the active fuel" which has been defined to be 344.5 inches above vessel tero.

These trip settings represent the indicated water level.

i High radiation levels in the main steam line tunnel above that due to the normal nitrogen-and oxygen radioactivit/ is an indicatiun af leaking fuel. A scram is initiated whenever_such radiat^on level exceeds three times normal background.

For the performance of a Hydrogen Water Chemistry pre-implementation test, the scram setpoint may be changeu based on a calculated value of the radiation level expected during the test.

Hydrogen addition will result in an approximate one-to five-fold increase in the nitrogen (N-16) activity in the steam'due to Tncreased N-16 carryover in the main steam. The purpose of this scrum is to reduce the source of such radiation to the extent necessary to limit the amount of radioactivity released due to gross fuel failure.

Discharge of excessive amounts of radioactivity to the 1

environs is prevanted by the air ejector offgas monitors which cause an isolation of the main condenser offgas line to the main stack.

The MSIV closure scram is set to scram wnan the isolation valves are 10% closed in 3 out of 4' lines. _This scram anticipates the pressure and_ flux transient which o ild occur when the valves close. Cy scramming at this setting, the resultant trans,ont is less severe than either the pressure or flux transient which would otherwise result.

A resctor mode switch is provided which actuates or bypasses the various scram functions appropriate to the particular plant operating status.

'The manual scram function is active in all modes, thus providing for a manual means of rapidly inserting control rods during all modes of reactor operation.

The APRM (High flux in Startup or Refuel) system provides protection against excessive power levels,and.short reactor periode in the startup and intermediate power-rangen.

The IRM system provides protection against short reactor periods in these ranges.

l A source rango monitor (SRM) system is also provided to supply additional neutron level informstion during startup but has no scram functions (reference paragraph 7.6.1.4 of the Updated FSAR).

fhus, the IRM and APRM are required in the " Refuel" and "Startup/ Hot StanGby"l modes.. In the power range the APRM system provides required protection (reference paragraph 7.6.1.7 of the-_ Updated FSAR).

Thus the IRM System is not required in the "Run" mode.

The APAM's cover only the power range.

l The IRM's and APRM's provide adequate coverage in the startup and intermediate range.

The control rod drive scram system is designed so-that all of the. water which_is

discharged from the reactor by-a scram-can be accommodated in the discharge piping.

The scram discharge volume accommodates in excess of 60 gallons of water and is the low point _in the piping.. No credit was taken for this volume in the design of the discharge piping as concerns the amount of water which must be accommodated daring a scram.

During. normal operation the discharge volume is empty; however, should it fill with water, the uater discharged to the piping from the reactor could not be accommodated.

which wauld~ result in slow scram times or partial control rod insertion.

70 preclude this occurrence, level switches have been provided in the instrument volume l.RTS-186h" 3.1-12 08/92

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3.2 " BASES In addition to reactor protection instrumentation which initiates a reactor scram, protective instrumentatior has been provided which initiates action to mitigate the consequencco of accidents which are beyond the operator's ability to control, or terminates operator errors before they result in serious consequences.

The objectives of tha Specificationu ares l

1.

To enoure the effectivenous of the protective instrumentation when required l

including periodo when portions of such eyotema are out of ocrvice for l

maintenance.

When necessary, one channel may be made inoperable for briot I

intervalo to conduct required functional tests and calibrations.

2.

To properibe the trip settings required to aneure adequate performance.

Some of the settings on the instrumentation that initiate or control core and

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containment cooling have tolerances explicit ly stated where the high and low values are both critical and may have a substantial effect on safety.

The setpoints of other instrumentation, whG e only the high or low end of the setting has a direct bearing on safet,, are chonen at a level away f rom t he normal operating range to prevent inadvertent actuation of the safety system involved and exposure to abnormal situations.

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The instrumentation which it..tiates primary system isolation is connected in a dual buu arrangement.

The trip level settings given for reactor water level represent the indicated water level.

The reactor water level trip settings are defined or described in " inches" I

above the top of active fuel.

The term top of active fuel, however, no longer has a preciso physical meaning since the length of the fuel pellet columns has changed over time from that of the initial cora load.

Since the basis of all safety analycec is the absolute level (inches above vessel zero) of the trip settings, the

" top of tho active fuel" has been arbitrarily defined to be 344.5 inches above vessel zero.

This definition is the same as that given by Figure 5.1-1 of the Updated FSAR for the initial core and maintains the consistency between the various level definitions given in the FSAR and the technical opecifications.

The low water level instrumentatian set to trip at 170" above the top of the active fuel closes all isolation valves except those in Groups 1, 6,

7 and 9 (see notes to Table 3.7-3 for isolation valve groups). Details of valve grouping and required closing times are given l'

+cif ication 3. 7.

For valves which isolate at this level tnis trip setting 16

' equate to provent uncovering the core in the case of a break in the largeot line assuming a 60 second valve closing time.

Beguired closing times are leso than this.

The low-low reactor water level instrumentation is set to trip when reacter water level is 119.5" above top of the active fuel.

This trip initiates the HPCI and RCIC and tripo the recirculation pumps.

The low-low-low reactor water level instrumentation is set to trip when the water level is 18.5" above the top of the

! active fuel.

This trip activates the remainder of the ECCS subsystems, closes Group 7 valves, closes Pain Steam Line Isolation Valves, Main Steam Drain Valves, Recire Sample Valves (Group 1) and starts the emergency diesel generatort.

There trip level nottings were chosen to be high enough to prevent spurious actuation but low

{ rnough to initiate ECCS operation and primary system loolation so that post accident cooling can be accompliched and the guidelines of 10 CFR 100 will not be exceeded.

I For large breaks up to the complete circumferential break of a 22-inch recirculation l line and with the trip cetting given above, ECCS initiation and primary system isolation are initiated in time to meet the above criteria.

Reference Sections 6.3 and 7.3 of the Updated FSAR.

l RTS-186A 3.2-43 0B/92

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