ML20247Q937

Proposed Tech Specs Re Instrumentation Setpoints

ML20247Q937
Person / Time
Site:	Fermi
Issue date:	07/24/1989
From:	DETROIT EDISON CO.
To:
Shared Package
ML20247Q934	List: ML20247Q937 NRC-89-0152, Application for Amend to License NPF-43,removing Inequality Indicators from Numerical Values in Trip Setpoint Column of Instrumentation Tech Specs
References
NUDOCS 8908070290
Download: ML20247Q937 (19)

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( POWER DISTRIBUTION LIMITS 3/4.2.2 APRM SETPOINTS 1

LIMITING CONDITION FOR OPERATION 3.2.2 The APRM flow biased neutron flux-high scram trip setpoint (S) and flow

. biased neutron flux-high control rod block trip setpoint (SRB) shall be established according to the following relationships: ,

Emmacanr mP SEMINT(emud ALLOWABLE VALUE 5 4-4. (0.66W + 52%)T $<(0.66W+54%)T Sg Q (0.66W + 42%)T 5dI(0.66W+45%)T where: S and 5 W = Loop,B are in percent recirculation flow asofaRATED percentage THERMAL POWER, of the loop recirculation

.. flow which 3roduces a rated core flow of 100 million 1bs/hr, at 100% of RATED THERMAL POWER T = Lowest value of the ratio of FRACTION OF RATED THERMAL POWER divided by the MAXINUM FRACTION OF LIMITING POWER DENSITY. T is applied only if lep than or equal to 1.0 -

APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equu to 2 2 of RATED THERMAL POWER.

ACTION:

With the APRM flow biased aeutron flux-high scram trip setpoint and/or the flow biased neutron flux-hi@ control rod block trip setpoint less conservative

'C.~ than the value shown in the Allowable Value column for S or SRB, as above determined, initiate corrective action within 15 minutes and adjust S and/or

'S RB to be consistent with the Trip Setpoint value* within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or reduce l

THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.2.2 The FRTP and the MFLPD for each class of fuel shall be determined, the

..' value of T calculated, and the most recent actual APRM flow biased neutron y flux-high scram and flow biased neutron flux-high control rod block trip setpoints verified to be within the above limits or adjusted, or.the APRM gain readings shall be verified as indicated below,* as required:

a. At laast ' once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,

b. Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and

c. Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating with MFLPD greater than or equal to FRTP.

d. The provisions of Specification 4.0.4 are not applicable.

4

• With MFLPD graater than the FRTP during power ascension up to 90% of RATED THERMALPOWER;ratherthanadjustingtheAPRMsetpoints,theAPRMgainmaybe adjusted such that APRM readings are greater than or equal to 100% times HFLPD, provided that the adjusted APRM reading does not exceed 100% of RATED THERMAL

( POWERandanoticeofadjustmentispostedonthereactorcontrolpanel.

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2.2 LIMITING SAFETY SYSTEM SETTINGS

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2.2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS .

The Reactor Protection System instrumentation setpoints specified in Table 2.2.1-1 are the values at which the reactor trips are set for each para-meter. The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their Safety Limits during normal operation and design basis anticipated operational occurrences and to assist in mitigating the consequences of accidents. Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drif t

- allowance assumed for each trip.in the safety analyses. /9dd 8eES # 5 E A7'

1. Intermediate Range Monitor, Neutron Flux - High The IRM system consists of 8 chambers, 4 in each of the reactor trip systems. The IRM is a 5 decade 10 range instrument. The trip setpoint of 120 divisions of scale is active in each of the 10 ranges. Thus as the IRM is ranged up to accommodate the increase in power level, the trip setpoint is also ranged up. The IRM instruments provide for overlap with both the APRM and SRM systems.

The most significant source of reactivity changes during the power increase is due to control rod withdrawal. In order to ensure that the IRM provides the required protection, a range of rod withdrawal accidents have been analyzeo. The results of these analyses are in Section 15B.4.1.2 of the FSAR. The most severe case involves an initial condition in which THERMAL POWER is at approximately 1% of RATED THERMAL POWER. Additional conserva-tism was taken in this analysis by assuming the IRM channel closest to the l control rod being withdrawn is bypassed. The results of this analysis show that the reactor is shutdown and peak power is limited to 21% of RATED THERMAL POWER with the peak fuel enthalpy well below the fuel failure thres-hold of 170 cal /gm. Based on this analysis, the IRM provides protection against local control rod errors and continuous withdrawal of control rods in sequence and provides backup protection for the APRM.

2. Average Power Range Monitor For operation at low pressure and low flow during STARTUP, the APRM scram setting of 15% of RATED THERMAL POWER provides adequate thermal margin between the setpoint and the Safety Limits. The margin accommodates the anticipated maneuvers associated with power plant startup. Effects of increasing pressure at zero or low void content are minor and cold water from sources available during startup is not much colder than that already in the system. Tempera-ture coefficients are small and control rod patterns are constrained by the RSCS and RWM. Of all the possible sources of reactivity input, uniform con-trol rod withdrawal is the most probable cause of significant power increase.

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f WW M , 'O A  : Selection of Trip Setpoints and Allowable Values is based.on methods .

Jl established by the General Electric Instrument Setpoint. Methodology as c described 'in General Electric document NEDC-31336.-

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1. Core CCFL pressure diffarendial - 1 psi . Incorporate the ,ssumption that flow from the bypass to lower plenum must overcome a 1 psi 1 pressure drop in core.

~2. Incorporate NRC pressure transfer assumption - The assumption used in the SAFE-REFLD00 pressure transfer when the pressure is increasing t> was changed.-

. A few of the changes affect the accident calculation irrespective of CCFL. These changes are listed below.

a. Input Chance

1. Break' Areas - The DBA break area was calculated more accurately.

b.- Model Change

2. ' Improved Radiation and Conduction Calculation - Incorporation of CHASTE 05 for heatup calculation.

. - A list of the significant plant input parameters to the loss-of-coolant accident analysis is presented in Bases Table B 3.2.1-1.

3/4.2.2 APRM SETPOINTS The fuel cladding integrity Safety Limits of Specification 2.1 were based on a power distribution which would yield the design LHGR at RATED THERMAL POWER. .The flow biased simulated thermal power-upscale scram setting.and flow biased simulated thermal power-upscale control rod block functions of the APRM instruments must be adjusted to ensure that the MCPR does .1ot become less than 1.06 or that 2,1% plastic strain does not occur in the degraded situation. The scram settings and rod block settings are adjusted in accordance with the formula in this specification when the combination of THERMAL POWER and MFLPD indicates a higher peaked power distribution to ensure that an LHGR transient would not be increased in the degraded condition. ,

Operofion with a trip act (en ec,n eo va to've thw its "fi tp  ;

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[ cancf the. Allowalte % fue os eg u a l to or feu inan fp e c4tf1 .\

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methob es tc, bin A e<f b y ff,e (yen e ea / E tee fric_ kteumenf Se troisf Me1hocioley as cfescriled .4 General E lec tr.'c c/oc u me <f 4/E Df. - 3 t M C FERMI - UNIT 2 B 3/4 2-2  ;

INSTRUMENTATION

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3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION This specification ensures the effectiveness of the instrumentation used to mitigate the consequences of accidents by prescribing the OPERABILITY trip setpoints and response times for isolation of the reactor systems. When necessary, one channel may be inoperable for brief intervals to conduct required surveillance. Some of the trip settings may have tolerances explicitly stated where both the high and low values are critical and may have a substantial effect o.i safety. The setpoints of other instrumentation, where only the high or low end of the setting has a direct bearing on safety, are established at a level away from the normal operating range to prevent inadvertent actuation of the systems involved.

Except for the MSIVs, the sa'fety analysis does not address individual sensor response times or the response times of the logic systems to which the sensors are connected. For D.C. operated valves, a 3 second delay is assumed

• before the valve starts to move. For A.C. operated valves, it is assumed that the A.C. power supply is lost and is restored by startup of the emergency diesel generators. In this event, a time of 13 seconds is assumed before the valve starts to move. In addition to the pipe break, the failure of the D.C. operated valve is assumed; thus the signal delay (censor response) is concurrent with the 10 second diesel startup. The safety analysis considers an allowable inventory loss in each case which in turn determines the valve speed in conjunction with the 10 second delay. It follows that checking the valve speeds and the 10 second time for emergency power establishment will establish -)

the response time for the isolation functions. However, to enhance overall system reliability and to monitor instrument channel response time trends, the isolation actuation instrumentation response time shall be measured and recorded as a part of the ISOLATION SYSTEM RESPONSE TIME.

Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.

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3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION

~ The emergency ccre cooling system actuation instrumentation is provided to initiate actions ta mitigate the consequences of accidents that are beyond the ability of the operator to control. This specification provides the OPERABILITY requirements, trip setpoints and response times that will ensure [

effectiveness of the systems to provide the design protection. Although the instruments are listed by system, in some cases the same instrument may be used to send the actuation signal to more than one system at the same time.

Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.

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-r selection'of Trip setpoints and Allowable values is based on methods established by the General Electric Instrtment Setpoint Metixx3 ology as described in General Electric document NEDC-31336. -

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.t mas 3/4.3.4 ATWS RECIRCULATION PUMP TR!p 5Vaipi ACTUATION INSTRUMENTATION The anticipated transient without scres (ATWS) recirculation pump trip system provides a means of limiting the consequences of the unlikely accurrence of a failure to scram during an anticipated transient. The response of the Llant neralto this Company Electric postulated event Topical falls Report within dated NE00-10M9 the envelope March 1971 of stu(y events in MEDD-34222, dated December 1979, and Appendix 158.d of the FSAR.

Deration with a trip set less conservative than its Trip 5etpoint but ~

within its specified Allowable Value is acceptable on the basis that the -

diffemace between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.

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3/4.3.5 REACTOR CORE 150LATION COOLING SYSTEM ACTUATION INSTRUMENTATION 1

The remeter core isolation cooling system actuation instrumentation is provided to initiate actions to assure adequate core coolinn in the event of l

reactor isolation from its primary heat sink and the loss o" feedwater flow to l

the reactor vessel without providing actuation of any of the emergency core j cooling equipment. ,

Operation with a trip set less conservative than its Trip 5etpoint but within its specified Allowable Value is acceptable on the basis that the

{ difference between each Trip 5etpoint and the Allowable Valus is equal to or less than the drift allowance assumed for each trip in the safety analyses.

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. 3/4.3.6 CONTROL R0p BLOCK INSTRUMENTATION, The control red block functions are provided consistent with the -

requirements of the specifications in Section 3/4.1.4, Control Rod pronram Centrols and Section 3/4.2 Power Distribution Limits. The trip logic es D>u t 3/4.3.7 SONITORING INSTRLMENTATigN -

3/4.3.7.1 RADIATION 9CNTTORING INSTRUMENTATION The OPERABILITY of the radiation monitoring instrumentation ensures that; (1) the radiation levels are continually asasured in the areas served by the individual channels; (2) the alem er autanatic action is initiated when the radiation level trip setpoint is exceeded; and (3) sufficient infomation is available on selected plant parameters to monitor and assess these variables following an accident. This capability is consistent with 10 CFR Part 50, Appendix A General Design Criteria 19, 41, 60, 61, 63, 64.

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MONITORING INSTRUMENTATION (Continued) 3/4.3.7.12 RADI0ACTIVEGij$EOUSEFFLUENTMONITORING-INSTRUMENTATION The radioactive gaseous effluent monitoring instrumentation is provided to monitor and control, as. applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gasw us effluents.

The alarm / trip'setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the 00CM utilizing the l-. system design flow rates as specified in the ODCH. This conservative method is used because the Femi 2 design does not include. flow rate measurement -

D devices. This will ensure the alarm / trip will occur prior to exceeding the limit:s of 10 CFR Part 20. This instrumentation also includes provisions for monitoring and controlling the concentrations of potentially explosive gas mixtures in the main condenser offgas treatment system. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design t,riteria 60,, fW and 64 of Appendix A to 10 CFR Part 50.

3/4.3.8 TURBINEbVERSPEEDPROT,ECTIONSYSTEM This specification is provided to ensure that the turbine overspeed protection system instrumentation and the turbine speed control valves are OPERABLE and will protect the turbine from excessive overspeed. Protection 1-from turbine excessive overspeed is not required to protect safety-related components, equipment, or structures. However, it is included in order to

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improve overall plant reliability.

3/4.3.9 FEEDWATER/ MAIN TURBINE TRIP SYSTEMS ACTUATION INSTRUMENTATION l

l- The feedwater/ main turbine trip systems actuation instrumentation is L provided to-frif tf ate action of the feedwater system / main turbine trip system in the event of a high reactor vessel water level due to failure of the feedwater controller under maximum demand.

Opera fin w ifA c< trip .s e+ (ess cou e r va tive th u ih Teip setpo14 but wi+4in ih spec wec/ 4/fowalole Va fu e & ,

accephble >s He bush that fhe c6He<eue be twe en cae(

Trip Getpoint and he R/lowable I/alue ls egual 4 oc les s ilum -ffte. drid allowarug cas a mec/ 6e enc 4 fe p a n he

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At% fy analyds . Selects.n of Te y, Mpo tA w<ct 8/fo+ bk I/duu is humf vn me %b e.st<<hb)Itecf by 6e Generei E(eenic .Tm r<ume.d 6fpo nf Mc ft<ocioloyy as clesulbed' ir, Generd Eleefde doc umed UEP(_ -3133t; .

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