ML19254E848

Revised Pages for 790830 Proposed Amend to License DPR-30 & Tech Specs App a

ML19254E848
Person / Time
Site:	Quad Cities
Issue date:	10/24/1979
From:	COMMONWEALTH EDISON CO.
To:
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ML19254E844	List: ML19254E843 ML19254E848
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NUDOCS 7911020397
Download: ML19254E848 (15)

Text

i i ATTACHMENT Revised Pages for Proposed License Amendment and Technical Specification Changen for Quad-Cities Unit 2 DPR-30 License 4

Page 4 Appendix A Page #

t 1.0-4 1.1/2.1-2 1.1/2.1-3 1.1/2.1-4 1.1/2.1-8 1.1/2.1-9 1.2/2.2-2

- 1.2/2.2-3 i 3.1/4.1-1 3.3/4.3-9

! 3.3/4.3-10 3.5/4.5-9 i

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DPR-30

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j Yhte itsense shall be deemed to contean an.: te evh!.e. to c'e eendit tone ope. t f ted in the followf as Creerst setea **go attona

T in 10 CFR Chap'er 1. Part 20, ses s ' en 10 14 of

s' art in .

Section 40.41.sf Part 40. sectane W.% and 10 i'A of'rert 50

' and section 1012 of Part 'O, le ewbiert to a t t appt li e'.te provietone at ehe Act and ts: the rule s, reautat s, no ea1 ordeee i of the e ewin t ee len now or hereaf ter in effect; e .. . t s e . t. jec t to

• i the additional condit ione spet tfled or tea ucparate.t he l e.w I, a. M e a t s'sen Power Lewel j

? Cassionwealth Edteon le authorised to operate Quadetttien

Unit 88o. 2 at power levela not in escess of 2511 megewet te (thermal).

B.

,, Techalcal s.pecfficatfon.s

. The Technli .nl Specf fications contained in stocendtces A and 8, as revised through Menenent 8io.14 7, are ie hereby incorporated in the license. The I tci nsee I shall operate the factitty in accor#tante with the Technical specIficatfont.

h'

. ,3. C Restr_f.ct iims f

Operation in the constdown mode is permitted to 40%

power. Should off-normal feedwater heating be

. ket ea a n t

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7 kCPROperating imit and calculated peak pressure E

for the worst :ase abnormal operating transient

., remain bounding for the new condition.

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h D. gu alizer_veJ ve_ Res.t r 3,etj en i

j Asn. 12 The valvre in the e.ru lizer piping bettveen the re -

4/21/75 circulation reactor operation. loops shall be closed at a l l t tine s du s i r.g E. $_ecurity Plan -

i The licensee shall maintain in effect and fully implement all Air. 148 provisions of the Comission-approved physical security plan, lk including amendments and changes made pursuant to the authority L

2/23'79 of 10 CFR 50.54(p). The approved security plan consists of

!I documents withheld from public disclosure pursuant to 10 CFR

{' } 2.790(d), referred to as Quad Cities Nuclear Power Station it r,. Units Nos.1 and 2 Physical Security Plan dated as follows:

fk Plaa - November 18, 1977

!' 'tevision 1 - May 19,1978 Revision 2 - May 27,1978 l' Revision 3 - July 28,1978 f

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4 t g QUAD-CITIF.S

,  ;*. DPR-30

.h Y. Shutdown . The reactor is in a shutdown condition when the reactor mode switch is in the Shutdown position and no core alterations are being performed.

I. Hot Shutdown means conditions as above, with reactor coolant temperature greater than 212' F.

2. Cold Shutdown means conditions as above. with reactor coolant temperature equal to or less than j, 212 ' F.

! Z. Simulated Automatic Actuation - Simulated automatic actuation means aj. plying a simulate (. signal to the sensor to actuate the circuit in question.

l1'

}[ BB. Tramition Boiling - Transition boiling means the boiling regime between nucleate and film bo ti Transition boiling is the regime in which both nucleate and film boiling occur i stermittently,with neither 11 y being completely stable.

tb ij CC. Cris? cal Power Ratio (CPR) - The critical power ratio is the ratio of that assembly power which causes i'

some point in the assembly to experience transition boiling to the assemb!y power at the reactor condition orinterese a calculated by application of the GEXL correlation (reference NEDO-10958).

h DD. Minimum Critical Power Ratio (MCPR) - Th. .inimum incore critical power r to corresponding to the most limiting fuel assembly in the core.

g EE. Surveillance Interva! - Each surveillance requirement shall be performed within the specified surveil-

!4, MiiF lance interval with:

5 a. A maximum allowable extension not to exceed 25% of the surveillance interval.

- b. A total maximum combined interval time for any 3 consecutive surveillance intervals not to exceed j, 3.25 times the specified surveillance interval.

i b

FF.

Fraction of Limiting Power Der sity (FLPD) - he fraction cf limiting power j

i density is the ratio of the linear heat generation rate (LHGR) existing at a given location to the desi n 5 LH3R for that bundle type.

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Maxi =ut Fracticn cf Limiting Fo er Eensity (MFLPD) - 2e maritu: fraction of limiting pcver der.sity is the highest value existing in the-core of the

>- fraction of limiting power density (FLPD). -

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, EH. Fraction of Rated Power (FRP) - he fraction of rated power is the ratio of y, core thernal power to rated thermal power of 2511 MWth.

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QUAD-CITIES DP't-30 Where:

D. Reactor Water lect Nbuidown Condition)

FRP = fraction of rated Whenever the reactor is in the shut- gg down condition witt. irradiated fuel (2511 MWt) in the reactor vessel, the water level shall not be less than that MFLPb = maximura fraction of limiting powc: dens-corresponding to 12 inches above the ity where the limit-l top of ti.e activo fuel

• when it is ing power density seated in the core.

for =ach bundle is the design linear

• Top of active fuel is defined to be heat generation rate 360 inches above vessel zero (See for that bu nd ie .

Bases 3.2).

The ratio of FRP/MFLPD shall be set equal to 1.0 unless the actu-al operating value is less tlin 1.0 in whica case the actual operating value will be used.

This adjustment may also be performed by increasing the APRM gain by the inverse ratio, MFLPD/FRP, which accomplishes the same degree of pro-tection as reducing the trip setting by PRP/MFLPD.

2. APlui Flux Scran Tr:p Setting (Re-fueling or Startup and Hot Standby Mode)

When the reactor mede switch is in the Refuel or Startup flot Sta,dby posi-tion, the APRM scram shal' be set at less than or equal to 15% of rated witron flux.

3. IRh ' lux Scran Trip Setting The IRM flux scram setting shall be set at less than or equal to 1:0/125 of full tie.

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 r; (.65V(p+ 43) l

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QUAD-CIT 2ES DPR-30 The definitions used above for che APitM scram trip apply. In the event of oper-ation with a maximum fraction limiting power density (MFLPD) greater than the fraction of rated power (FRP), the se.tting shall be modified as follows:

? FRP

' S 6 (.65Wp + 43) MFLPD i 1 The definitions used above for the APRM l, scram trip apply, g

h The ratio of FRP to MFLPD shall be set equal to 1.0 unless the actual operat.ng l value is less than 1.0, in which case i

the actual operating value will be used.

This may also be perf$rmed by increasing the APRM gain by the inverse ratio, MFLPD/FRP, j

which accomplishes the same degree of pro-tection as reducing the trip setting by 1 -

P FRP/MFLPD.

\-! C. Reactor low water level scram setting i

I shall be 144 inches above the top of the 1

active fuel

• at normal e jerating condi-

' tions.

4 D. Reactor low water level ECCS initiation shall be 84 inches (+4 inches / 0 inch) lt above the top of the active fuel

• at 1

normal operating conditions.

E. TurM :e stop valve scram shall be s 10% valve closure from full open.

L I F. Turbine control valve fast closure scram shall (3' initiate upon actuation of the fast closure sole-p, noid valves which trip the turbine control 1.

valves.

G. Main steamline isolation valve closure scram shall be s 10% valve closure from full open.

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' H. Main steamline low- ~ . essure initiation of main steamline isolation val <e eksure shall be jj .

2 850 psig.

i of ac-tive fuel is defined to

• beTop 3 60 inches above vessel sero ill (See Bases 3 2)

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QUAD-CITIES DPR .,0 1

1.1 SAFETY LIMIT BASIS

• l The fuel claddireg integrity limit is set such that no calculated fuel darssge would occur as a resu1* of an abnormal operational transient. Because fuel damage is not directly observabic, a step-back approach is used to establish a safety limit such that the minimum critical power ratio (MCPR) is no less than the fuel l cladding integrity safety limit r.CPR > the fuel cladding integrity safety limit represents a cor.servative I margin relative to the conditions required to maint ain fuul clodding integrity.

The fuel cladding is one of the physical barricrn which separ ate radioactivo materials froat the cTrite s.

The '.ntegrity of t! cladding barrier is related to its sciative freedom from perforations or crack t r.g.

Alt . ugh some cort mon or use-related cracking may occur during the life of the cladding, fissien prcduct migrs* * ..,m this source is incrementally c?nulative and continuously reasurable. ruel cladd tr.g pe r-fora ,, however, can result from thermal stresses which occur frota reactor operation significantly above deetw conditions and the protection system safety rettings. While fission produce m - + ' fro 3 cl uding perforation is just as asaasurable as that from use-related cracking, the thermally caused cladda:: ca r n c-stions signal a threshold beyond which still greater thermal stresses may cause gross rather thar 1r. m eent-l al cladding deterioratien. Therefore, the fuel cladding safcty limit in de fined with margan to the c:7.di-l tions which would produce onset of transition boiltng (:tCFA of 1.0). Thcse conditions represent a signiti.

cant desertere frcra the condition ir. ten:' ' by design for planned operation. Therefore , the fuel clad.* : no L _,___ integrity esfety lim i s (stablished syi.a that no calculated fuel damage is expected to,0 r a s', I l a result Of anahmormal operational transtant. Basis.of the values derived for this apfatY 181 888 each fuel type is documer.ted in Reference 1 jj d A. Reactor Pra nare > 800 peig a.nd Core Flow > IC% of Stated t

Cnset of 2ransition boiling results in a decrease in heat transfer from the cladding and therefore

'[p. ." elevated cladding tempereture and the possibility of cladding failuro. However, tno ex1 rte nce of

• eritical power, or boiling transition is not a directly observable parmeter in an operet =7 r e act-A or. Therefore, the margin to boiling transition is calculated freu plant ope ra t ing par: sters such as core power, core flow, feedwater teoperature, and core power distribut ton. The cargin is: -ret i fuel assembly is characterized by the critical power ratio (CPR), which is tha ratio of P: b/ . ::

t power which would produce onnot of transition boiling dividei by the actual imCle poAct . -m

! minimum valma of this ratio for any bundle in the core is thu minimum critical power retto (.m).

It is assumed that the plant operation is controlled to the nominal protective setponts v a tra instrumented varia'bles (rigure 2.1-3). <

, The MCPR fuel cladding integrity safety limit has suf ficient con:.ervatism to assure that in the avend of an abnormal operational transient initiated from the rormal operating condition, more tnan 99.9%

of the fuel rods in the cora are expected to avoid boiling transition. The margin between ECpR of l 1.0 (onset of transition boiling) and the safety limit, is derived from a detailed statisttcal {

t analysis considering all of the uncertainties in monitoring the core operating state, including i uncertainty in the boiling transition correlation (ses e.g.. Roference 1). Because the boilir.g

r transition correlation is based on a large quantity of full-scale data, there is a very high con-fidence that operation of a fuel assembly at the condition of MCPR = tho fuel claddir,g integrity l safety limit would not produce boiling transicion. I Hm

n r r, if butling transitiori were to occur, cledding perforation would not be expected. Cladding temperatures would increase to approximstely 1100 F, which is below the perforation temparature of the cladding material. This has been verified by tests in the General Electric Test Reactor (C ETR) ,

where similar fuel operated above the critical heat flux.for a significant period of time (3C min-utes) without cladding perforation.

I If reactor pressure should evor exceed 1400 psia during nornal power operation (the limit of I ,

applicability of the boiling transition correlation), it would be assured that the fuel cladding integrity safety limit has been violated.

In addition to the boiling transition limit (MCPR) operation is constrained to a maximum L H CP.s 17.5 kw/ft for 7 x 7 fuc1 and 13.4kw/ft for all 8x8 fuel types. This constraint is established by l specification 3.5.3 to strain for abnormakrovide adequate safety. margin to 1% plastic operating transients initiated from high power conditions. Specification 2.1.A.1 provides for equivalent safety margin for transients initiated from* lower power con-ditions by adjusting the APRM flow-biased scram setting by the I ratio of FRP/MFLPD. l 1.1/2.1-4 1252 020 . - .

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QUAD-CITIES DPR-3 0-I; An increas. in the APRM scram trip setting would decrease the margin present before the

,j fuel cladding integrity safety limit. is reached. The APRM scram trip setting was determined by an analysis r' margins required t'o provide a reasonable range for maneuvering during

{ a, operation. Reducing this operating margin would increase the frequency of spJrious SCI 8ms, which have an adverse ef fect on reactor safety because of the resulting therr.al stresses.

f' Thus, the APRM scram trip setting was selected because it provides adequate margin for the

t. 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 IRGR transient peak is not increased for any combination 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 MFLPD is greater than the fraction of rated power (FRP) .

1 The adjustment may be accomplished by increasing the APRM gain bV the reciprocal

• 1 of'FRP/MFLPD. This provides the same degree of protection as reducing. , ,

the trip setting by FRP/MFLPD by raising the initial APRM readings

closer to the trip settings such that a scram would be received at the same point in a transient as if the trip settings had been re-

! duced by FRP f MFLPD*

l

2. APRM Fisx Scram Trip Setting (Refuel or Startup/ Rot Standby Mode) 3 For operation in the Startup m-de while the reactor is at low pressure, the APRM scram setting i.

of 15% of rated power provides adequate thermal margin between the setpoint and the safety

'! limit, 25% of rated. The margin is adequate to acco==odate anticipated maneuvers associated with power plant startup. Effects of increastnq pressure at zero or lew void conten ara minor, cold witter frcn vorm m'l 51c .iuring startup 1J not much COlddr W h 16 t,ulfddy in khc system, temperature cocrficienu are small, and control red patterns are constrained to be

. uniform by operating procedures backed up by the rod worta minimiter. Of all possible sources t

of reactivity input, uniform control rod withdrawal is the most probable cause of significant l power rise. Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks, and because several rods must be moved to change powe. by a signifi-cant percentage of rated power, the rate of power rise is very slow. Generally, the heat flux

['

is in near equilibrium with tha fission rate. In an assumed uniform rod withdrawal approach to the scram level, the rate of power rise is no mo.ce than 5% of rated power per minute, and the APRM symtem would be more than adequatb to assure a scram before the power could exceed the safety limit. The 15% AFRM scram remains active until the mode switch is placed in the Run position. This cwitch oe:urs when reactor pressure is greater than 850 psig, k 3. IRM Flux Scram Trip Setting 1

I i

The IRM system consists of eight chambers, four in each of the reactor protection system logic '

4 channels. The IRM is a 5-decade instrument which covers 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 setting would be 120 divisions for that range r likewise, it the instrument were on Range 5, the scram would be 120 divisions on that range.

Thus, as the IRM is ranged up to accorranodate the increase in power level, the scram trip set-ting is also ranged up.

The most significant sources of reactivity change during the power increase are due to control i rod withdrawl. In order to ensure that the IRM provides adequate protection against the single rod withdrawal error, a range of rod withdrawal accidents was analyzed. This analysis included starting the accident at various power levels. The most severe case involves an initial condi" ion in which the reactor is just suberitical and the IRM system is not yet on scale.

! Additional conservatism was taken in this analysis by assuming that the IRM channel clo3e Ft t ,

the withdrawn rod is bypassed. The results of this analysis show that the reactor is scramed i and peak power limited to 1% of rated power, thus maintaining MCPR above the fuel cladding l integrity safety limit. Based on the above analysis, the IRM provides protection against local control rod withdrawal errors and centinuous withdrawal of control rods in sequence and

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provides backup protection for the APPM. ,

L 1252 021 1.1/2.1-8 l

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B. APRM Rod Block Trip Setting Reactor power level may be varied by moving control rods or by varying the recirculatien flow rate. The APRM system provides a control rod block to prevent gross rW 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 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 substantial 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 relationships therefore the worst-casc MCPB which could occur during steady-state operat ion is at 108% of rated thermal power because of the APRM rod block trip setting. The actual power distribution in the core is established by specified control rod sequences and is monitored continuously by the incore LPRM system. As with APRM scram trip setting, the APRM rod block trip setting i,s adjusted downvard if the maximum fraction cf limit- '

ing power density exceeds the fraction of rated power, thus preserving the APRM rod block safety margin. As with the scram setting, this may h ccomplished by adjusting the APRM gains.

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 becausa the level instrumentation is density compensated.

D. Reactor Low Low Water Level ECCS Initiation Trip Point The emergency core cooling subsystems are d 41gned to provide sufficient cooling to the ore to dissipate the energy associated withthe loss-of-coolant accident and to lidit fuci NJding temperature to well below the cladding melting temperature to assure that core geometry z w ains intact and to limit any cladding metal-water reaction to less than 1 To acc< aplish ti str intended function, the capacity of each emergency core cooling system component was establisud based on the reactor low water level scram setpoint. To lower the setpoint of the low water level scram would increase the capacity requirement for each of the ECCS components. Thus, the reactor vessel low water level scram was set low enough to permit margin for oc aration, yet will not be set lower because of ECCS capacity requirements.

l; The design of the ECCS components to meet the above criteria was dependent on three previously sat parameters: the maximum break sire, the low water level scram setpoint, and the ZCCS initiation setpoint. To lower the setpoint for initiation of the ECCS could leac'. to a lcas of 4

effective core cooling. To raise the ECCS initiation setpoint would be in a safe direction, but it would reduce the margin established .o prevent actuation of the ECCS during normal operation or during normally expected transients.

I I

E. Turbine Stop valve Scram The turbine stop valve closure serei trip anticipates tNe pressura, neutron flux, and heat flux increase that could result from rapid closure of the tuabine etop valves. With a scram trip setting of IC; of valve closurc from full open, the resultant increase in surface heat flux is limited the such that worst-case MCPR remains transient abovethe that assumes theturbine MCPP.bypass fuel cladding integrity safety limit even during j is closed.

. F. Turbine Control Valve Fast Closure Scram The turbine control valve fast closure scram is provided to anticipate the rapid increase in pressure and neutron flux resulting from f ast closure of the turbine control valves due to a load rejection and subsecuent failure of the bypass, i.e., it prevents MCPR from beco::iin.g less than the MCPR fuel cladding integrity safety limit for this transient. For the load rejection without bypass transient from 100% power, the peak heat flux (and therefore LHGR) increascs on the order cf 15% which provides wide margin to the value corresponding to 1% plastic strain of the cladding.

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1.2 SAFETY I IMIT IIASES f

The reactor coolant system integrity is an important ' arrier in the prevention of uncontrolled release of fission products. It is essentis hat the integrity of this system be protected by establishing a pressure limit to be observed

} for all operating cono.oons and wheneser ther, it irradiated fuel in the reactor vessel.

, ne pressure safety limit of 1325 psig as measured by the vessel steam space pressure indicator is equivalent to 1375 psig at the lowest elevation of the reactor coolant sptem. The 1375 psig value is denved from the design pressures of the reactor pressure sessel and coolant system piping. The respective design pressures are 1250 psig at 575* F and 1875 p<ig at 560' F. The pre-sure safety liruit was chosen as the lower of the pressure transi:nts perrnitted by the applicable design coJes ASMI Uo:ler anJ Pressure Vessel Code Section !!! for the pressure set.el, j and USASI B31.1 Code for the reactor coolant spiern piping The ASME Botter and Pressure Vessel Code permits

pressure transients up to 10% over design prosure (110% x 1150 = 1375 psigl. and the USAS! Code permits t pressure transients up to 20% over the design pressure (120% a 1175 = 1410 psig). The safety limit pressure of 1375 psig is referenced to the lowest elevation of the primary coolant system. Evaluation methodology used ressure is not exceeded fortoany assure reloed that is, documented this safety lituit in Regerence 1.

~

ne design basis for the reactor pressure vessel males evident the substantial margin of protection against failure at the safety pressure limit of 1375 psig. The vessel has been designed for a Eeneral membrane stre:s no greater than 26,700 psi at an mternal pressure of 1250 psig; this is a factor of 1.5 below the yield strength of 40.100 psi at 575

• F. At the pressure limit of 1375 psig, the general membrane st;ess will only be 29,400 psi, still safely below the yield strength.

l' The relationships of stress inels to yield strength are comparable for the prirnary system piping and provide a similar margin of protection at the established safety pressure Lmit.

! The normal operating pressure of the reactor coolant system is 1030 psig. For the turbine trip orless of efectricalload transients, the turbine trip scram or generator load rejection s: ram together with the turbice bypass syste : limits the pressure to approximately 1100 psig (References 2,3 and4). In addition. rressure relief valveshave been provided to

. , reduce the probabthey of the safety calves oper.itng in the event that the turb:ne bypass shculd fa~

Finally, the safety vahes are sired to keep the reactor coolaa; system pressur below 2

1375 psig with no credit taken for relief sa!ves dunng the postulated full closure of all MSIVs without direct (salve position switch) scrani. Credit is taken for the neutron flax scram, however.

l' The indirect flux scram and safety vahe a:tuation, _ . provide adequat: margin

~ below the peak allowable vessel p. assure of 1375 psig.

Reactor pressure is continuously monitored in the control room during operadon on a 1500 psi full-scale pressure recorder.

1  :

} References

,

• 1. " Generic Reload Fuel Application", NEDE-240ll-P-A*

. 2. SAR, Section 11.22 l  % 3. Quad Cities 1 Nuclear Power Station first reload license f submittal, Section 6.2.4.2, February 1974.

4. GE Topical Report Nr.DO-20693, General Electric Boiling Water Reactor No. 1 liccnaing submittal for Quad Cities Nuclear Power Station Unit 2, Decorabor 1974 Approved revision number at timo relodd analyscs are perforraed.

1.2/2.2-2 1252 023

- QUAD CITIES 2.2 LIMITING SAFETY SYSTEM SETTING BASES In compliance with Section III of the ASME Code, the safety valves must be set tc open at no higher than 103% of design pressure, and they must limit the reactor pressure to no more than 110% of design pressure. Both the high neutron flux scram and safety valve actuation are required to prevent overpressurizing the reactor pressure vessel and thus exceeding the pressure safety limit. The pressure scram is available as backup protection to the high flux scram. Analyses are performed as described in the

" Generic Reload Fuel Applicatiom " NEDE-240ll-P-A (approved revision number at time reload analyses are performed) for each reload to assure

~t h'at the pressure safety limit is not exceeded. If the high-flux scram were to fail, a high-pressure scram would occur at 1060 psig.

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h 3.1/4.1 IIEACTOlt PRO'IECTION SYSTEM IJMI'llNG CONDITIONS FOR Orr.R ATION SURVEll.l.ANCE RFQtilRDIENTS

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Applicability: Appliemidh:3 :

Applie> to the antirumentation ai.J ateli.iicit d4 Apphes to the wrveillance of the instrumentation um which initiate a reartnr wram. and aswciaicJ devices .nich initiate reactor scra m.

Objecth e: Objecth e:

T. auure the operahility of the reactiu proveetion To speciiy the type e.nd frequency nf surveillance in sptem. he applied in the protection inurumentation, i

SPECIFICATIONS

4. The setroints. minimam number of trip sys. A. Instrumentation systems shall be functior. ally scm>. and minimum number of instrument tested and calibrated as indicated in Tables j channels that .nust be operable for each posi- 41 1 and 4.12 rttpettisely.

1 M. tion of the teattor mode switch shall he as 8 -

W given in Tahles 3.1 1 through 3.1 4. The system B. D.iity during reactos power operation. the mre

response times from the openmg of the sensor nower distribution shall be checked for maximum contact up to and including the opening of the fraction of limiting power dens-l ,. trip actuator contacts shall nnt exceed 100 ity (MFLPD) and compared with the i Us58 "d - fraction of rated power (FRP) i B* If, during operatica, the maximum j l when operating above 25% rated i i fraction of limiting power dens- thermal power.

ity exceeds the fraction of rated power when operating above 25%

C. When it i> determined that Sannst is failed

! rated thermal power, either:

an the unsafe cendita n arid Cc..amn I of Ta-

1. the APRM scram and rod ble: 3.I I through 3.13 cannot be met that block settings shall be trip system must be put in the tripped condit.on immediaiety. Au ther RPS ch.:nnels that mon-reduced to the valuer

. stor the : eme variable shall N ft.ni- w!!v given by the ceruntions ,,.edwim R W i6 m wm . e d

, in Cpecifis t :c.m 2 . l . cs .1 fai cil th umet nu) be untopp':J far a period et l

cr.d 2.1.B. h ic tie, also time not to esteed I hour in conduct this be accomplished by testing. A.s Icing as the trip system with the increasing the !JIU4 failed thannel contains at least one operable

' gain as described. channel n.onitoring that same variabte. that

. therein. trip syuem niay he pl. iced in the untripped positioit for short periods of time to .illow functional testing of all RPS instrument chan.

nets as specified by Taale 4.1-1.The trip system may be in the untripp d position for no more ihm 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> per funct.or.al teu perital for tht,

"I

2. the pcuar listribution d ill bc chm;;cd : u ;h t that the maximum fraction of limiting power density ')" -D nu lo:iger exer crla thr: \

fraction of ra t<.d povier.

3.1/4.1-1

Q' I.,50g QUAD CITIES g .Q E, DPR-30 b, the delayed neutron fraction etiosen for the bounding reactivity curve l

c. a beginning-of-life Dopoler reactivity feedback

d. scram times slower than the Technical Specification rod scram inJertion rate (Section 3 3.c.1)

e. the maximum possible rod drop velocity of 3 11 fps

f. the design accident and scram reactivity shape function, and

g. the moderator temperature at which criticality occurs In most cases the worth of insequence rods or rod segments in conjunction l with the actual values of the other important accident analysis parameters described above, would most likely result in a peak fuel enthalpy sub-stantially less than 2D cal /g design limit. l Should a contro'. drop accident result in a peak fuel energy content of 280 cal /g. fewer than 660 (7 x '

7) fuel rods are conservatively estimated to perforate. This would result in an c:Tsite dose well below the guideline value of 10 CFR 100. For 8 x 8 fuel, fewer than 850 rods are conservatively estimated to perforate, with nearly the same consequences as for the 7 x 7 fuel case because of the rod power differences.

TL ;od worth minimizer provides automatic supervision to assure that out of sequence control rods will not be withdrawn or inserted; i.e., it limits operator deviations from planned withdrawas sequences (reference SAR Section 7.9). It serves as a backup to procedural control of control rod worth. In the event that the rod worth minimizer is out of service when required, a licensed operator or other qualified technical employee can manually fulfill the control rod pattern conformance function of the rod worth minimizer. In this case, the normal procedural controls are backed up by independent procedural controls to assure conformance.

4. The source range monitor (SRM) system performs no automatic safety system function,i.e., it has no scrsm function. It does provide the operator with a visual indication of neutron level. This is needed for knowledgeable and etheient reactor stattup at low neutron levels. The ccinsequences of reactivity accidents are functions of the irdtial neutron flux.The requirement of at least 3 counts per second assures that any transient, should it occur, begins at or above the initial value of 104 of rated power used in the analyses of transients from cold conditions. One operable SRM channel would be adequate to mom.or ine approach to criticality using homogeneous patterns of scattered control rod withdrawal. A minimum of two operable SRM's is provided as an added conservatism.

5. The rod block monitor (RBM) is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power operation. Two enannels are provided, and one of these may be bypassed from the console for maintenance and/or testing. Tripping ofone of the channels will block erroneous rod withdu al soon enougn to prevent fuel damage. T1.is system backs up the operator, who withdraws control rods according to a written sequence. The specified restrictions with one channel out of service consewatively assure tnat fuel damage will not occur due to rod withdtr.wal errors when this condition exists. During reactor _

operation with certain limiting control rod patterns, the withdrawal of a designated single control rod could result in one of more fuel rods with MCPR's less than the MCPR fuel claddin6 inte6rity safety limit.During use of such patterni it isjudged that testing of the RBM system to assure its operability prior to withdrawal of such rods will assure that improper withdrawal does not occur. It is the responsibility of the Nuclear E . deer to identify these limiting patterns and the designated rods either when t.e , * -

established or as they develop due to the cecarrense cfiwpr Me cu .:rui rm : 1 m ar r -

~'

patterns.

1252 026 3 34 .3-9

e QUAD CITIES

} -g Scram Insertion Times C.

The control rod system is analyzed to bring the reactor suberitical at a rate fast enough to prevent fuel damage, i.e., to prevent the .':CPE from becoming less than the fuel cladding integrity safety limit. l 2

Analysis of the limiting power transient shows that the negative reactivity rates resulting from the scram wit,h tSe average respc:.se :f all the drives as given in the above specification, provide tne require.

proLecticn, and MCPR remains greater than the fuel cladding integrity safety limit. The minimum amount of reactivity to be inserted during a scram is controlled by permitting no more than 10h of the operable rods to have long scram times. In the analytical treatment of the transients,390 milliseconds are a!! owed between a neutron st.isor reaching 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 270 milliseconds. Approp imately 70 milliseconds after neutron flux reaches the trip point, the pilot scram valve solenoid deenergizes. Approximately 200 milliseconds later, control rod motion begins.The time to deenergize the pilot valve scram solenoids is measured during the calibration tests required by Specification 4.1. The 200 milliseconds are included in the allowable scram insertien 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 fo!!owing a <

representative sample of control rods will be scram tested shutdown.

Scram times of new drives are approximately 2.5 to 3 seconds; lower rates of change in scram times following init' plant operation at power are expected. The test schedule ((

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

!. The history of drive performance accumulated to date indicates that the 90% insertion times ornew and overhauled drives approximate a normal distribution about the mean which tends to become skewed 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 measurement of the e, cram performance of the drives surrounding a drive excceding the expected rang of scram perferms.; M A : m.. . s. :: it.. 40 pros iJe assuranse that local scram time limits are i.a: ew;eJtJ. Conanued momtormg of othci drives exceeding the expected range of scram times provides surveillance of possib!c anomalous performance.

The numerical values assigned to the predicted scrara performance are based on the analysis of the Drceden 2 startup da'a and of data from other HWR's such as Nine Mile Point and Oyster Creek.

The n.; :mc; eof- . I nt s -

. averare, shmaid .w '

as an indication ut a viematic pa,bkm wan ucunei i o.1 . .  % if.the nura.,er (a diis es

. . We rods.

exhibiting such scram times cwtds eight the abwablc nui..i 3.3 / 4. 't- 10}

QUAD-CITIES P00RORL3 Mot . O cycle by y . ;p that uater can be run throg!' the drain lino and actuating the air operated vahes by operation of the following sensors:

1) loss of air

2) equipment drain < ump high level

3) vault high level

d. The condenser pit 5-foot trip cir-cuits for each channel shall be checked once a morth. A logic system functional te: shall be per-formed curing each refueling outage.

I. Average Planar LIIGR I. Average Planar LliGR During steady-state power operation, the average Daily during steady state operatior. linear heat generation rate (APLilGR) of all the a,bove 2'5fo rated therma 1 power, rods in any fuel assembly,as a function of average - the ayerage planar LHGR shall planar exposure, at any axial location, shall not be determined, f exceed the maximum average planar LIIGR O shew t# ris=< 3.5-> . ir i #v tt-- durir.g operation it is determined by n_rmal sur-i a veillance that the limiting value for APLilGR is J. Local LifGP. ! being exceeded, action shall be initiated witi. 15 minutes to restore operation to .iithin the pre-Daily during steady-state power nperation above 25% of rated thermal power. the local scribed limits. If the APLiiGR is not returned in within the prescribed limits within 2 hours, the LHGR shall be determined. i1 l reactor shall be brought to the cold shutdown lt condition within 36 hours. Survedlance and correspondmg action shall cor.tinue until reactor operation is within the presenhed limits.' J. Local LilGR

,              During steady-state power operation, the linear heat generation rrte (LilGR) of any rod in any fuel assembly at any axial location shall not exceed the maximum allowable Ll!GR.                       l If at any time dunng operation it is determined by narmal surveillance that the hmitmg value for LilGR is being exceeded, action shall t.s initiated within 15
.              mir.utes to restore operation to within the pre-scribad limits. If the LilGR is not returned to O                                                                                                           1252 028 3.5 / 4. 5-9

                       '1     %I                     QUAD CITIES

{i 00 0 15.1 l DPR-30 within the prescribed limits within 2 hours, the reactor shall be brought to the cold shutdown condition within 36 hours Surveillance and cor-responding action shall continue entil reactor operation is within the prescribed limits. bhximum nllcWable L'!GR for all 8XC fuel types is 13.4 KW/ft. For 7X7 and mined oxide fuel, the maximum allowable LHGF, is as follows: LHG R,,,,, < LilG R , 1 -( S P/ P ),,,,( L/ l., ) w here: Lif G R, = design LilGR

                        =     17.5 kW/ft.

( A P/ P ),,,, = maximum power spiking pen.nlty

                        =     .035 initial core fuel
                        =    .029 reload 1, 7 x 7 fuel

, h = .028 reload 1,7x7mixedoxidefuel(l

L, = total core length

                        =    12 feet I           L            =    Axial distance from bottom of core 1

K. Minimum Critical Power Ratio (MCPR) K. Minimum Critical Powor Ratin (MCPR) During steady state operation MCPR shall be The MCPR shall be determined daily during greater than or equal to steady-state power operation abo' e IP,' of 1.35 (7 x 7 fuel) rated thermal power. 1.35 (8 x 8 fuel) at rated power and flow. If at any time during . operation it is determmed by normal sursedlance that the limiting value for MCPR is bemg esceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed hmits. If the steady state MCPR is not returncJ to within the prescr: bed knuts w ohin 2 hours, the re.ic:or , sh21 be h:.n.. c ta 'hs c M sna. .ow n t .'i en witiun 36 haai>. hane.'iaxe aaa correr Jmg action sha'l continue ur'til reactor opera: ton is within the frewribed huat s. For core ih.w s other than rated, th::.c nannnal valoca of M(1% dull 1259._ h be increaud by a factur of kg where kg is as shown in Figure 3 5 2. 15 / .8.b lo}}