ML19344D159
| ML19344D159 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 03/07/1980 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML19344D157 | List: |
| References | |
| NUDOCS 8003110437 | |
| Download: ML19344D159 (6) | |
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ENCLOSURE QUAD CITIES UNIT 2
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Revised Proposed Technical Specification Pages:
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QUAD-CITIES DPR-30 D.
Reactor Water Lesel (Shutdown Condition) where 5
whonever the reactor is in the shut-down condition with irradiated fuel t
a p er in the reactor vessel, tho water (2511 MWt) level shall not be less than that corresponding to 12 inches above the NFLPD = maximum-fraction of l
top of the active fuel
- when it is limiting powcr dens-seated in the core.
ity where the limit-ing power density
- Top of active fuel is defined to be for each bundle is 360 inches above vessel zero (See the design linear Bases 3.2).
heat generation rate for that bundle.
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The ratio of FRP/MFLPD shall be set equal to 1.0 unless the actu-al operating value is less than 1.0 in which case the actual operating value will be used.
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- 2. APRM Flux Scran Trip Setting (Re-fueling or Startup and Hot Standby Mode)
When the reacto; mode switch is in the Refuel or Startup Hot Standby posi-tion, the APRM scram shall be set at less than or equal to 15% of rated neutron flux.
The IRM flux scram setting sha!! be set at less than or equal to 120/125 of full scale.
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- 4. When the reactor mode switch is in the startup or run position, the reactor shall not be operated in the natural circula-tion flow mode.
B.
APRM Rod Block Setting The APRM rod block setting shall be as shown in Figure 2.1-1 and shall be:
S s (.65V/p+ 43) l 1.1/2.1-2
QUAD-CITIES DPR-30 The definitions used above for the APM scram trip apply. In the event of oper-ation with a maximum fraction limiting power density (l'.FLPD) greater than the fraction of rated power (FRP), the setting shall be modified as follows:
BE S.4 (.65WD + 43)
MFLPD The definitions used above for the APM scram trip apply.
The ratio of TRP to MFLPD shall be set equal to 1.0 unless the actual operating value is less than 1.0, in which case the.2ctual operating value will be used.
l C. menetor low wat er level scram setting shall be 144 inches above the top of the active fuel
- at normal operating condi-
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tions.
D. meneter low water level recs initiation shall be 84 inches (+4 inches /-0 inch) above the top of the active fuel
- at l.
O nozinal operating conditions.
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'E. Turbine stop valve scram shall be s 10% valve L
closure ffom full open.
I F. Turbine control valve fast closure scram shall initiate upon actuation of the fast closure sole-noia valves which trip the turbin: control l
valves.
e O. Main steamline isolation valve closure scram f
31.11 bas 10% valve closure from full open.
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. E. Main steamline low. pressure initiation of main steamline isolation valve closure shall be a 850 psig.
- Top 60 inches above vessel zeroof ac-be 3 (See Bases 3 2)
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DPR-30' Aa incrrco in tha APM scrtm trip setting would decrease tha margin procent b:f tro the fuel cladding integrity safety limit, is reached. The APPM scram trip setting was determined r
.by an analysis of margins required to provide a reasonaole range for maneuvering during operation. Reducing this operating margin would increase the frequency of spurtous scrams, which have an adverse ef fect on reactor safety because of the resulting ther.al stresses.
Thus, the APPM scram trip setting was selected torause it provides adequate margin for the fuel cladding integrity safety limit yet allows operating margin that reduces the possibil-ity of unnecessary scrams.
The scram trip setting must be adjusted to ensure that the LEGR transient peak is not increased for any cor.bination of maximum fraction of limiting power density (MFLPD) and reactor core thermal power. The scram setting is adjusted in accordance with the formula in Specification 2.1. A.1, when the MFI.PD is greater than the fraction of rated power (FRP).
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2.
AFRM Flux Scram Trip Setting (Refuel or Startup/ Rot Standby Mode)
For operation in the Startup node while the reactor is at low pressure, the APRM scram setting of 15% of rated power provides adequate thermal margin between the setpoint and the safety 11 alt, 25". of rated. The margin is adecuate to acec=odate anticipated maneuvers associated with power plant -tartup.
Effects of increasing pressure a zsro or Icw void content are minor, cold water frc:. c:ar::: m1LM Juring startup 10 not much coldar 1, nan that already in the system, temperature c:af hetan:s 4:e small, and control red patterns are constrained to be e
uniform by operating procedures bt:xed up by the rod worth minimizer. Of all possible sources of reactivity input, uniform control red withdrawal is the most probable cause of significant power rise. Because the flux distrioution associated with uniform rod withdrawals does not involve high local peaks, and because several rods must be moved to change power by a signifi-cant percentage of rated power, the rate of power rise is very slow. Generally, the heat flux le in near equilibrium with the fission rate.
In an assumed uniform rod withdrawal approach
' I to the scram level, the rate of power rise is no more than 5% of rated peer per minute, and the APRM system would be more than adecuat'e to assure a scram before the power could exceed the safety limit. The 15% AFM scram remains active until the mode switch is placed in the Run position. This switch occurs when reactor pressure is greater than 850 psig.
3.
IBM Flux Scram Trip Setting-The IRM system consists of eight chambers, four in each of the reactor protection system logic channels. The IRM is a 5-decade instrument which cuvers the range of power level between.that i
covered by the SRM and the APRM. The 5 decades are broken down into 10 ranges, each being one-half a decade in size.
The IRM scram trip setting of 120 divisions is active in each range of the IRM.
For example, if the instrument were on Range 1, the scram settin7 would be 120 divisions for that ranger -
likewise, if the instrument were on Range 5, the scram would be 120 divisions on that range.
Thus, as the IRM is ranged up to accommodate the increase in power level, the scram trip set-ting is also ranged up.
The most significant sources of reactivity change during ths power increase are due to control rod withdre.wl. In order to ensure that the IBM provides adecuate protection against the eingle rod withdrawal error, a range of red withdrawal accidents was analyzed. This analysis included starting the accident at various power levels. The most severe case involves an initial condition in which the reactor is Just suberitical and the I M system is not yet on scale.
Additional conservatism was taken in this analysis b/ assuming that the IM channelC10Se st t the witMer n red s t;:2ssed. The re M t: :f enn snalysts show that the reacter is scra=cc l a.d p sk pact limited to IL of rated Oc..cr, thus maintaining P.0PR absve the fuel cladding integrity safety limit.
Based on the above anslysis, the IRM provides protection against local control rod withdrawal errors and continuous withdrawal of control rods in sequence and providas backup protecticn for the APM.
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p QUAD-CITIES
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DPR-30 B.
APRM Rod Block Trip Setting Reactor power level may be varied by moving control rods or by varying the recirculation flow rate. The APRM system provides a control rod block to prevent gross rod withdrawal at constant recirculation flow rate to protect against grossly exceeding the MCPR Fuel Cladding Integrity Safety Limit. This rod block trip setting, which is automatically varied with recirculation l
loop flow rate, prevents an increase in the reactor power level to excessive values due to control rod withdrawal. The flow variable trip setting provides substantf al margin from fuel damage, assuming a steady-state operation at the trip setting, over the entire recirculation flow range. The margin to the safety limit increases as the flow decreases for the specified trip setting versus flow relationshipt therefore the worst-case MCPR which could occur during steady-state operation is at 100"'. of rated thermal power because of the APRM rod block trip eetting. The actual power distribution in the core is established by specified control rod sequences and is monitored continussly by the incore LPRM system. As with APRM scram trip setting, the APRM rod block trip setting is adjusted downward if the maximum fraction of limit-ing power density exceeds the fraction of rated power, thus preserving the APRM rod bicck safety margin.
C.
Reactor Low Water Level Scram The reactor low water level scram is set at a point which will assure that the water level used in the bases for the safety limit is maintained. The scram setpoint is based on normal operat-ing temperature and pressure conditions because the level instrumentatior. is density cu:pensated D.
Reactor Low Low Water Level ECCS Initiation Trip Point The emergency core cooling subsystems are designed to provide suf ficient cooling to the core to dissipate the energy associated withthe less-of-coolant accident and to limit fuel cladding temperature to well below the :ladding melting temperature to assure that core geometry remains intact and to limit any cladding metal-water reaction to less than 1%.
To accomplish their intended function, the capacity of each emergency core cooling system component was established based on the reactor low water level scram setpoint. To lower the setpoint of the low water level scram would increase the capacity recuirement for each of the ECCS components. Thus, the reactor vessel low water level scram was set low enough to permit margin for operation, yet will not be set lower because of ECCS capacity requirements.
The design of the ECCS comoonents to meet the above criteria was dependent on three previously
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set parameters: the maxi =um break size, the low water level scram setpoint, and the ECCS ini ation setpoint. To lower the setooint for initiation of the ECCS could lead to a logs of.
effective core cooling. To raise the ECCS initiation setpoint would be in a safe direction, but it would reduce the margin established to prevent actuation of the ECCS during normal operation or during normally expected transients.
B.
Turbine Stcp valve Scram The turbine stop valve closure scram trip anticipates the pressure, neutron flux, and heat flux increase that could result from rapid closure of the turbine stop valves.
With a scram trip setting of IC*; of valve closure
- rom full open, the resultant increase in surface heat flux is limited such that MCPR remains above the.MCsa fuel cladding integrity safety limit even during l
the worst-case transient that assumes the turbine bypass is closed.
F.
Turbine control Valve Fast closure Scram The turbine control valve fast closure scram is provided to anticipate the rapid increa.e in pressure and neutron flux resulting from fast closure of the turbine control valves aue to a load rejection and subsecuent failure o' the bypass, i.e.. it prevents MCPR from becoming less than the MCPR fuel clidding integrity safety limit for this transient. For the load rejection without bypass transient from 100% power, the peak heat flux (and therefore LHGR) increases on the order of 15% which provides wide margin to the value corresponding to 1% plastic strain of the cladding.
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QtfAll-Cllil:S 1)PH..to 3.1/4.1 IIEACTOR P!(OTECTION SYSTEM IJMillNG CONI)lTIONS FOR Orr. RATION SURYl{ll.l.ANCE : (. QUIRE \\lENTS Applieability:
Applieshihf 3 i
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Applie, to the inurumentatten anJ.is.r ;iarnt d -
Arrhes to the surveillance of the instrumentation s ces s hich inili.ite a rcartrir seram.
and asswiated devites which initiate reactor scrJ rt' 06]ecthe:
Objecche:
To awure the operahility of the reaooi pro eetion To specify the type and frequency of rarveillante to e
sy.sem, he applied to the proiettion instrumentation.
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SPF.CIFICATIONS i
A.
'the setpoints, minimum number of trip sys-A.
Instrumentation systems shall be functior. ally scms, and minimum numbar of instrument tested and calibrated as indicated in Tables channels that must be operrble for each posi.
41 1 and 412 ccipedisely.
l A
tion of she reauot mode switch shall he as 1
Not given in Tames 3.1 1 through J.1 4. Ttw syvem B.
D. ily during reauni power oyration, the core response times from the opening of the sensor power distribution shall tectiecked for maximum I
contact up to and iratudin; the opening of the fraction of limiting power dens-j trip actuator contaus shall not execed 100 ity (MFLPD) and compared with the
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fraction of rated power (FRP)
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B* If, during operation, the maximum when operating above 25% rated a
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fraction of limiting power dens-thermal power.
ity exceeds the fraction of rated power when operating above 25%
When :. is Jeterminyd that a channelis failed I-rated thermal power, either:
In the unsafe conditaan and Column I of Ta.
block settings shall be trip system must te put in the inpped enndition reduced to the values jmmediately. All other RPS ch.:nnels th.at mon-stor the same variab!e shail b: fanuu.na!!v given by the ecuations teued v.it!.in A t..m Th
.n, nsu:r. w nh !1-
-l in Specificati.cau 2.1..i.1 fai;cd th.innel n iy b: untripp':d'for a period or and 2.1.3.
time not to csceed I hour to conduct this testing. As Ions as the trip system with the faileJ thannel contains at le.ist one operahic channel rnonitorip; that same variable. that trip system may be g.laceJ in the untripped position for short periods of time to allow functional testing of all RPS instrument chan-nels as specif;cd by Tatte l.1 1.The trip system may be in the untripped position for no more than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> per functior.al test peririt for thin 2.
the pcwer distribution
"I "I' d.*ll Lv cW.. :W :,ucit that the ma:timum fraction of limiting power density D**g - ))'
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'3 Y no-longer excr:cds the i#*
fraction of rat *.d poucr.
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