ML20126K542

Proposed Tech Specs,Correcting Discrepancy Between Reactor Protection Sys Delay & Response Times

ML20126K542
Person / Time
Site:	Quad Cities
Issue date:	05/12/1981
From:	COMMONWEALTH EDISON CO.
To:
Shared Package
ML20126K540	List: ML20126K537 ML20126K542
References
NUDOCS 8105180253
Download: ML20126K542 (7)

Text

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O Enclosure 1 Quad Cities Station Unit 2 Proposed Technical Specification Changes i

Revised Pages: 1.1/2.1-2 1.1/2.1-2a 1.1/2.1-8 .

1.1/2.1-9 3.1/4.1-1 ,

3.3/4.3-10 l i

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8105180J53

QUAD-CITIES- ,

DPR-30 7

where:

D. Resetor Water led (Shutdown Condialen) ,

i FRP = fraction of rated Whenever the reactor is in the shut- * * .

dcun condition with irradiated f uel f'#

in the reactor vessel, the water  !

MLPD = maximum fraction of level shall not be less than that limiting power dens.

corresponding to 12 inches e.bove the ity where the limit-top of the active fuel

• when at is ,

ing power density ,

seated in the core, for each bundle is the design linear  ;

• Top of active fuel is defined to be heat generation rate 360 inches above vessel sero (see for that bundle.

Bases 3.2). '

The ratio of TRP/rELPD shall be set equal to 1.0 unless the actu-I al operating value is less than 1.0 in which case the actual operating value will be used. f This adjustment may also le performed by increasing the APIN gain by the

! inverse rStio, MFLPD/FRP, which l acecrplishes the same degree of pro-I tection as reducirv3 the trip setting

' by TRP/MFuv.

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2. APRM Flu Scram Trip Setting (Re.

fueling or Startup and liot Standby i Mode) ,

When the reactor mode switch is in the r Refuel or Startup flot Standbv posi.

tion, the APRM scram shall be ut at '

ku than or equal to IST, of rated neutron flux.

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3. IRM Flux Scram Trip Setting The IRM flux scram witing sha!!be ut at f kas than or equal to 120/125 of full scale.

4. When the reactor mode switch is in the startup or run position, the reactor shall not be operated in the natural cucula.

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:

l S 6 (0.58WD + 50) j 1.1/2.1-2

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• QUltD-CITIES  ;

DPR-30 . .

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. The definitions used above for the APRM ocram trip apply. In the event of oper-ation with a maxi. mum fraction limiting '

power density (MrLPD) greater than the '

fraction of rated power (FRP), the f.ett ing r(ill be n.odified as folio.as s ,

, S 6 (0.58WD + 50)

.

• The definitions used above for the APRM i cerers trip apply. .

The ratio of rRP to MrLro shall be set eque.1 to 1.0 unless the actual operating value is less than 1.0 in which case the actual operating value will be used.

This adjustment may also be performed

. . by increasing the APTN gain by the inverse ratio, MFLPD/PRP, which acecrTalishes the same degree of pro-tection as reducing the trip setting i

. by TRP/MPLrV.

C. Reactor Icu water level scram setting shall be 144 inches above the top of the activo fuel

• at normal operating ec,ndi-  :

tions.

. D. Reactor low water level tecs initiation shall be 84 inches (44 inches /-0 inch) above the top of the active fuela at

• normal operating conditions.

E. Turbine stop valve scram shall be I 10% valve closure from full op:n. ,

F. Turbine control valve fast closure scram shall

. Initiale upon actuation of the fast closure sols.

noid valves which trip the turbine centrol valves. ,

O. Main steamline isolation valve closure scram shall be s 10% valve closure ftom full open.

H. Main steamline low pressure initiation ofinsin e . $leamline isolation valve closure shall be k s25 psig. .

• Top of active fuel is defined to be 360 inches above vessel zero ,

(See Bases 3 2) ,

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1.1/2.1-2a -

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DPR-30'* ,

An increase in the APM ceren trip setting would dscrosso the margin present

(.y fuel cladding integrity safety limit le reach *d, j by an operation.

analysis of margins required to provide a1 reasonable stresses.  ;

which have an adverse ef fect on reactor safety because of the resulting therra the I Thus, the APM scram trip setting was selected because it ~

ity of unnecessary scrams. i Tha scram trip setting sust be adjusted to' ensure that the ImR transient peak is not .

t Ancreased for any c wnbination of sarinum fraction of limiting power density (M?LP The scram setting is adjusted in accordance with'the forrula reactor core thermal pe.er.

• in specification 2.1.A.1, whqn the Mr1.Pp is greatsr than the (raction of rated i f

We adjustment may be acecmplished by increasing the APfH gain by the reciprocal of FRPAfLPD. h i prcnides the same degree of protection as reducing the trip '

setting by FRP/MPLFD by raisirq the initial APfN readings closer to the trip settirgs  !

such that a scram would be recieved at the same point in a transient as if the trip b settings had been reduced by PRP My .

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2. APM Flux Scram Trip setting (Rafuel or Startup/ dot Standby Hode)  !

por operation in the Startup snede while the resetor is at acw pressure, the APM sl of 15% of rated pe.<er grovides adequate thermal nargin between limit, 25% of rated. f with oower plant startup.

Effects of increasing r,rosaure at sero or low void cetntent are ,

ninor, cold water from sources ave 11st>1e during startcp is not euch colder thun lj '

system, temperature coefficients are small, andTrotyol god patterns are secrees constraine of all possible aniform by operating piecedures backed up by the red worth minimiser.  ;

of reactivity input, uniform control rod withdrawal is the most probable cause of significa!

power rise. Secause the flux distribution associated with uniforra red withdrawa involve high local peats, and because seversi rods must be moved to change j

Generally,.the heat pcuer, flux <

cant percentage of rated power, the rate of power rise is very slow.In an assumed u.niform

~

rod!

is in near equilibrium with the fission rate. d  ;

• to the scram level, the rate of poser rise.is no more than 5% of rated power per minute, d an the A?RM system would be core than adequat's to assure a scram be

1. the Run safety limit.Thi's switch occurs when reactor pressure is greater than $25 psig.

posit' ion. *

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3. IRM Flux Scram Trip Setting [

The charmels.

1M system consists of eight char.bers, four la occh of thel The $ decades are broken down into 10 ranges, each being c

rovere6 by the SRM and the APM. . . f one-half a decade in oise.

• For example,

• j The IRM scram trip setting of 120 divisions is metive in each range of the IM.  ;

if the ir>strument were on Range 1, the scram setting would be 120 divirions for that ranger!

likewise, if the instrumer.t were on Range 5, the scraa would be 120 divisic,ns on that rani Thus, as the IM is ranged up to acconenodste the increase in power level, the scres ting is also ranced up. .

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The most significan*, sources of resetivity change duriin the pow'st increase are jd

. In order to ensure that the 2M provides odeguate protection against This ana'the lysis -

red withdrev1.-

single red withdraval error, a range of rod withirawal accidents was analysed.The mo!f included starting the accident at various power levels.Amitial condition] '

t to !

ocale. ,

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• Additional conservatism was taken la kis analysis by assuming Jing that l the withdrawn rod is bypassed.

and peak poser limited Based toon 15theofaboveratedanalysis, power,the thus ARMmaintaining provides protection McPR above it against

!ategrity' safety limit.

local control rod withdrawat errors and continuoas withdrawal of control r[

provides backup protection for the APM.

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• l QUAD-CITIES

= . . DPR-30 ,

l( 0. AFM Rod Block Trip setting Reactor power level may be varied by moving control rods or by varying the recirculatien flow l rate. The APM system provides a control rod block to prevent gross rod withdrawal at constant recirculation flow rate to protect against grossly exceeding the MCPR Puel Cladding Integrity '

Safety Limit. This rod block trip setting, which is automatically varied with recirculation loop facu rate, prevents an increase in the reactor pcuer level to excessive values due to *

'cootrol rod withdrawal. The flow variable trip setting provides substantial margin frem fuel entire recirculatien damage, assuming a steady-state operation at the trip setting, over the '

flow range. The margin to the safety limit increases as the flow decreases for the specified trip setting versus flow relationships therefore the worst-case MCPR which could occur during .

I steady-state operation is at 10F% of rated thermal power because of the APM rod block trip setting. The actual power distribution in the core is established by specified control red As with APRM scram trip sequences and is monitored continuously by the incore LPM system.  !

setting, the APRM rod block trip setting is adjusted downward if the r.axi.=um fraction of li=it- j Ang power density exceeds the fraction of rated power, thus preserving the APRM rod bloch safety mergin. As with the scram setting, this may be accceplished by adjusting the APIN gains. l C. Ratetor Low water Level Scram The reactor low water level scram is set at a point which will assure that the water level use' r in the bases for the saf ety limit is n;aintained. The scram serpoint is based on ncrral operat-ing temper,ature and pressure conditions because the level instrumentat Aon is density eccpensated.

D. Basetor Low Lew Water Level ECCS Initiation Trip Point r' sufficient cooling to the core The energency core cooling subsystems are designed to provide to dissipate the energy associated withthe loss-of-coolant accident and to licit fuel clacding to assure that core geometry remains temperature to well below the cladding melting temperature To accorplish their frr intact and to limit any cladding metal-water reaction to less than 1%. was estaclashed

di . intended function, the espacity of each emergency core cooling system cc .ponent

• based on the reactor low water level scram setpoint'. for To eachlower theECCs cf the setpoint of the lowThus, co. penents. water the level seras would increase the capacity reovirementlow enough to permit margin for operation, yet will [

reactor vessel low water level scram was set not be set lower because of ECCS capacity requirements.

The design of the ECCS cceponents to meet the above criteria was dependent on three and the previously ICCS set parameters: the maxiecm break size, the low water level scram setpoint, initiation setpoint. To lower the setpoin*. for initiation of the ECCS could lead to a loss cf .

2 effective core cooling. To raise the ECCS initiation setpoint would be in a saf e direeston, but it would reduce the r.argin established to prevent actuation of the ECCS during normal l operation or during normally expected transients.

E. Turbisa stop Valve scram The turbine stop valve closure scram trip anticipates the pressure, neutrcn flux, and heat flux increase that could result from rapid closure of the turbine stop valves. With a acram trip .

setting of 10" of valve closuro from full open, the resultant increase in surface heet flux is '

limited such that MCPR remains above the MCPP. fuel cladding integrity safety lamit even during the worst-case transient that assumes the turbine bypass is closed. .

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F. Mrbine control Valve rast closure Scram The turbine control valve f ast closure scram is provided to anticipate the rapid increase in l pressure and neutron flux resulting from f ast closurei.e., of the turbine control valves due to a it prevents MCPR from bgccming less [

load rejection and subsecuent failure of the bypass,  :

than the MCPR fuel cladding integrity safety le.it for this transient. For the the peak 2c'ac heat flux rejection w?thout bypass transient from 100% power,15% which provides wide increases on the order of (and therefore LP.GR) corresponding margin to the value to 1% plastic strain of the cladding.

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t 3.1/4.1 IlEACTOlt P)(OTECTION SYSTEM  !

IJMillNG CONI)lTIONsi FOR OPr.II ATlON SURWil.I .A NCI: 110Q1 filtr.41r.NTS

Apphenbility  !

Applicahihrp Applici no the inurument.ition anJ atte dateil J.9 Applies to the surveillance of the inutumentation tres w hich initiate a reartnr wram.

and amwiancJ devices which in4tiate resetor eeta m.

Objeciher OWecchei 1.s awre the operability of tlw ceaoot prowtion To specify the type and frequer ey m'surveillante in

3. tem.

tw applied to the protection inurumentauon.

i SPF.CIFICATIONS A. Tbc icipoints. minimum number or trip sys. A.

tems, and minimum number of instrument Inurument.uien tystemt shall he functier. ally thannch that roust be operchie for each pnti. tested and calibrated e indicaird in Tables 41 1 and 4.14 respecthcly.

sion of the reactor anode swinh shall be as C given in Tables 31 1 ihtough J.14. The syuem respon*r times from the opening of the semor B. Daily durit g reaetrv power tyeration. the core sintact up to and including she opening of the

• poset distrihuuur shaltis checkeJ for maximum l erip actuator contaus shall nas exceed 50 fraction of limiting power dens.

smiliseconds ity (MFLPD) and compared with the If, during operation, the maximurr. fraction of rated power (FRP)

• fraction of limiting power dens- when operating above 25% rated ity exceeds the fraction of rated thermal power.

power when operating above 25% .

rated thermal power, either: .

p it in detesnu.gd that a channelis fadeJ in the un are cendinan and Column 1 of Ta.

1. the APRM scram and rod Wes 3.11 through 3.14 tunnnt be met, that block settings shall be l'I synem P mun be pin in thr inpped tundition reduced to the values iminedialcly. Alloch e RPS ch.:nnels that mon.

given by the equations llor the same variable shall be functirinally l

in Specifications 2.1.A.1 neued within R hours The trip systern *ith the

• and 2.1.D. This may also failed eh.innel may be untripped for a period of '

be accomplished by -

time not to emnt i bour to coneun ihit erstint. ^5 knr at the tri increasing the APRM gain I'I'"' '*'""'I ***I" "' p system s-ith the 3'd" ""' 'P"* hI' as described therein. N"'I '"*"I'"'I"I 'h*' ""'N 'h

trip systern may be placed ".in'he t unt ripped t position for short periods of time to allou '

functinnal sesiing of all RpS inurument chan.

ach as specified by Table 4.l.l.The trip tystem misy be in the untripped potirion far no mnre '

than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> per functior.al tesi perical for this

2. the power distribution ""I"l' l , shn11 be changed such '

( that the maximum fraction l of limiting power density  ;

no longer execeds the fraction of rate:d power. >

I 3.1/4.1-1

.o mes & Mn

I gvnu VAeAuv DPR-30 C. Scram' Insertion Times '

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The control rod system is analyzed to bring the reactor suberitical at l I a rate fast'enough to prevent fuel damage, i.e., to prevent the MOPR from becoming less than the fuel cladding inte5rity safety limit.

power, transient shows ' that the negative

~Knalysisofthe"limitin$ngfromthescramwiththeaverageresponseof reactivity rates' result

' all the drives as given in the above specification, provide the required ,

protection, and MCPR remains greater than the fuel cladding integrity ,

safety limit.

The minimu.m amount of reactivity - i to be inserted during a scram is controlled by permitting no more than 10% of the operable rods to have long scram times. In the analytical treatment of the transients,290 milliseconds are allowed between a l neutron sensor re' a ching the scram point and the start of motion of the control rods. This is adequate and conservative when compared to the typically observed time delay of about 210 milliseconds. Approx.

imately 90 milliseconds after neutron flux reaches the trip point, the pilot scram valve solenoid doenergizes and 120 milliseconds later the control red motion is estimated to actually begin. However, 200 milliseconds rather than 120 milliseconds is conservatively assumed for this time interval in the  !

transient analyses and is also included in the allcwable scram insertion times specified in Specification -

3.3.C. .

The scram times for all control rods will be determined at the time of each refueling outage. A representative sample of control rods will be scram tested following a l shutdown. l Scram times of new drives are approximately 2.5 to 3 seconds; lower rates of change in scram times  ;

The test schedule

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following initial plant operation at power are expected.

' provides reasonable assurance of detection of slow drives belbre system deterioration beyond the limits of Specification 3.3.C.The program was developed on the basis of the statistical approach outlined below

( and judgment. - '

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.' The history of drive performance accumulated to date indicates that the 90% insertion times of new and  ;

I overhauled drives approximate a normal distribution about the mean which tends to become ske'wed  ;

toward longer scram times as operating time is accumulated. The probability of a drive not exceeding the )

mean 90% insertion time by 0.75 seconds is greater than 0.999 for a normal distribution. The j measurement of the scram performance of the drives surrounding a drive exceeding the expected range i of scram performance will detect local variatior.s and also provide assurance that local scram time limits  ;

are not exceeded. Continued monitoring of other drives exceeding the expected range of scram times ,

provides survei!!ance of possible anomalous performance.

l The numerical values assigned to the predicted scram performance are based on the analysis of the i Dresden 2 startup data and of data from other HWrs such as Nine Mile Point and Oyster Creek.  !

('..' The occurrence of scram times within the limits, but significantly longer than average, should be viewed as an indication of a systematic problem with control soJ drives, especially if the number of drives

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exhibiting such scram times cacceds eight, the allowable number of inoperable rods. j 3.3M.3-10 i 6 6 h