ML20126C342

Amend 51 to License DPR-30,changing Tech Specs to Support Util Review of Future Reloads & Modifying Condition 3.C to Assure Conservative Max Critical Power Ratio Operating Limit During Coastdown Operation

ML20126C342
Person / Time
Site:	Quad Cities
Issue date:	03/20/1980
From:	Ippolito T Office of Nuclear Reactor Regulation
To:
Shared Package
ML20126C340	List: ML20126C337 ML20126C342 ML20126C346 ML20126C350
References
NUDOCS 8003310189
Download: ML20126C342 (51)

Text

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. a'og[o~'g UNITED STATES 2 >W e NUCLEAR REGULATORY COMMISSION h " ' $ WASHINGTON, D. C. 20555 t J

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COMMONWEALTH EDISON COMPANY AND TOWA-ILLIN0IS GAS AND ELECTRIC COMPANY DOCKET NO. 50-265 QUAD CITIES UNIT NO. 2 A'iENDMENT TO FACILITY OPERATING LICENSE Amendment No. 51 License No. DPR-30

1. The Nuclear Regulatory Commission (the Comission) has found that:

A. The application for amendment by the Comonwealth Edison Company l (the licensee) dated August 3,1979, as supplemented on October 24, i 1979 and March 7,1980, complies with the standards and require-  !

ments of the Atomic Energy Act of 1954, as amended (the Act), and i the Commission's rules and regulations set forth in 10 CFR Chapter I ,

B. The facility will operate in conformity with the application, the )

provisions of the Act, and the rules and regulations of the  !

Commission; ll C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2. Accord'ingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment and paragraph 3.B and 3.C of Facility License No. DPR-30 are hereby amended to read as follows:

8003310 h%-

-2' 3.B Technical Specifications The Technical Specifics ~. ions contained.in Appendices A and B, as revised through Amendment No. 51 , are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.C Restrictions Operation in the coastdown mode is permitted to 40% power.

Should off-normal feedwater heating be necessary for extended periods during coastdown (i .e. , greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) the Licensee shall perform a safety evaluation to determine if the MCPR Operating Limit and calculated peak pressure for the worst case abnormal operating transient remain bounding for the new condition.

3. This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

% /'

• Thomas W f

<Aycsc.t 4 4Chief

. /ppolito, Operating Reactors Branch #3 Division of Operating Reactors

Attachment:

Changes to the Technical Specifications Date of Issuance: March 20, 1980

[

ATTACHMENT TO LICENSE AMENDMENT NO. 51 FACILITY OPERATING LICENSE NO. DPR-30 DOCKET NO. 50-265 i

1. Remove the following pages & insert identically numbered pages:

1.0-2 3.3/4.3-3 1.0-4 3.3/4.3-4 1.1/2.1 -1 3.3/4.3-8 1.1/2.1-2 3.3/4.3-9 1.1/2.1 -3 3.3/4.3-10 1.1/2.1-4 3.3/4.3-11 1.1/2.1 -5 3.4/4.4-3 1.1/2.1-6 3.5/4.5-9 1.1/2.1 -7 3.5/4.5-10 1.1/2.1-8 3.5/4.5-13 1.1/ 2.1 -9 3.5/4.5-14 1.1/2.1-10 3.5/4.5-14a 1.1/2.1-11 Fig. 3.5-1 (2 sheets) 1.2/2.2-1 3.5/4.5-17 1.2/2.2-2 3.6/4.6-4 1.2/2.2-3 3.1/4.1 -1 3.1/4.1 -3 3.1/4.1 -5 3.1/ 4.1 -7 ,.

3.2/4.2-5 3.2/4.2-6 3.2/4.2-7 3.2/4.2-8

3.2/4.2-11 3.2/4.2-12 3.2/4.2-14 3.2/4.2-15

2. Figure 2.1-2 is deleted.

3. Figure 3.5-1 is being replaced by 4 pages.

i 1

4

I QUAD-CITIES DPR-30 J H. Limiting Conditions for Operation (l.CO) -The limiting conditions for operation specify the minimum acceptable levels of system performance necessary to assure safe startup and operation of the fa When these conditions are met, the plant can be operated safely and abnormal situations can be safely controlled.

1.

Limiting Safety Setem Setting (LSSS) The limiting safety system settings are settings on instrumenta.

tion which initiate the automatic protective action at a level such that the safety limits will not he exceeded. The region between the safety limit and these settings represents margin, with normal operation lying below these settings. The margin has been established so that with proper opera the ' instrumentation, the safety limits will never be exceeded.

K.- Logie System Functional Test A logie system functional test means a test of all relays and contacts a logic circuit from sensor to activated device to ensure all components are operable per design inte Where possible, action will go to completion; i.e., pumps will be started and valves opened.

2 L. Modes of Operation A reactor mode switch selects the proper interlocking for the operating or shutdown condition of the plant. Following are the modes and interlocks provided:

1. Shuidown In this position, a reactor scram is initiated. power to the control rod drive 3 is rence L and the reactor protection trip systems have been deenergized for 10 seconds prior to petmWre iN manual reset.

2. Refuel - In this position, interlocks are established so that one control rod only may be wi'ndrawn when flux amplifiers are set at the proper sensitivity level and the refueling crane is not over the reactor. Also, the trips from the turbine control valves, turbine stop valves, main steam isolition valves, and condenser vacuum are bypassed. If the refueling crane is over the reactor, all rod, must be fully inserted and none can be withdrawn.

3. Startup/ Hot Standby In this position,the reactor protection scram uip2. initiated by condenser law vacuum and main steamline isolation valve closure, are bypassed, the low pre. use main steamle isolation valve closure trip is bypassed,and the reactor proicction system is energized, with IRM and APRM neutron monitoring system trips and control rod withdrawal interlocks in service.

4. Run In this position the reactor system pressure is at or above 850 psig, and the reactor protection system is energized.with APRM protection and RMB interlocks in service (excluding the 159 hig flux scram).

M. Operable A system or component shall be considered operable when it is capable of performing its

. intended function in its required manner.

K Operating Operating means that a system or component is performing its intended functions in its required manner.

O. Operating Cycle Interval between the end of one refueling outage for a particular unit and the end of the next subsequent refueling outage for the same unit.

P. Prhnary Containment Integrity - Primary containment integrity means that the drywell and pressure suppression chamber are intact and all of the following conditions are satisfied:

.l. All' manual containment isolation valves on lines connecting to the reactor coolant system or containment which are not required to be open during accident conditions are closed.

51 Lo-2 Amendment No.

, , , -+ ,. -- t

QUAD-CITIES DPit-30 Y.

Shutdown . The reactor is in a shutdown condition when the reactor mode switch is in the Shutdow n 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 me. ins conditions as ahos e, with reactor coolant temperature equal to or less than 212 F.

2.

Simulated Automatie Actuation . Simulated .iutomatic actuation means applyin;; a simulated signal to the sensor to actuate the circuit in question.

88. Transition Railing Transition boiling means the boiling regime between nucleate and film boiling.

Transition bmhng is the regime in which both nucleate and film boiling occur intermittently, with neither type being completely stable.

CC. Critical Power Ratio (CPR) . The critical power ratio is the ratio of that assenthly power which causes some point in the assembly to experience transition boiling to the assembly power at the reactor condition ofinterest as calculated by application of the GEXL correlation (reference NEDO 10958).

DD. Min.imum Critical Power Ratio (MCPR) - The minimum incore critical puwer ratio corresponding to the most limiting fuel anembly in the core.

EE Suncillance Intenal . Each surveillance requirement shall be performed within the specified surseil.

lance interval with:

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 3.25 times the specified surveillance interval. i FF.

Fraction of Limiting Power Density (FLPD) - The fraction of limiting pcVer density is the ratio of the linear heat generation rate (LHGR) existing at a given location to the design LHGR for that bundle type.

GG.

Maximum Fraction of Limiting Power Density (MFLPD) - The maximum fraction of limiting power density is the highest value existing in the. core of the fraction of limiting power density (FLPD). -

HH.

• Fraction of Rated Power (TRP) - The fraction of rated power is the ratio of core themal power to rated thermal power of 2511 MWth.

Amendment No. 51 1.0-4

QllA D-CITilS DPR-30 1.1/2.1 l'UI'L CLADDING IN11:GRITY SA FETY l.lMIT 1.lMillNG SArt.TY SYS10.M Sr.TilNG Applicability: ApplienWiit):

The safety I;mits established to preserve the fuel The limiting sarmy system senings apply to trip cladding integrity apply in these variables which settings of the imtruments and devices whith are monitor t.ie fuel thermal Lehnior. provided to piescnt the fact claddin; integrity safety limits fro;n be.r.g ( uceJed.

Objetlite: Objectiie:

The ob;cetive of the safety limits is to establish The objective 01 ine hmiting sa rety systern settings hmits below which the integrity of the fuct el.idding is to defme the les el of the proteo variab:es at s bch is preserved. automatic protective action is initiated to prevent the fuel eladdirg integrity safery limits troin M ny exceeded.

SPECirICATIONS l A. Rescior Prt:uuse > S00 psig and Cure riow A. Neutron riux Trip Settins

> 10% of Rated The existence of a minimum critical The limiting safety systern trip settings shall be power ratio (McPR) less than 1.00 as specif.ed be!ow.

f or core loadir.q petterna cont oin-ing no retrofit Bxfl fuel (two water I. APRM l' lux Scram Tr#p Settmp (Run l rods) or 1.07 for core lu.eding Mode) I pattern 5 containing retrofit 6x8 l fuel shall constits.te violstion of When the reactnr mode swit:h is in ll:e t he. fuel c14.dding integrity safety Run position. el.e APitM tM saam j limit, '

sett'or shall be as show n in h tme 2.1 1 and shall be- l B. Core Therrnal Power Limit (Reactor Pressure '

ts 800 psi;:) SMN+W D l When the reactor pressure is c S00 psig or with a m?ximum setpoint of 120?e for core flow a less than 10"o of rated, the core core now equal to 96 x 10' lb/hr and thermal power shall not exceed 25% of rated I*#

thermal power. where:

C. Power Trar.sient S = seuing in pneent i rated power

!. The neutron Out shall not exceed the g, g

(,g g, g,g g;, , gcm,,,, g scram setting established in Speabea- , ngg tson 2.1.A for longer the.n 1.5 seconds f }ow of % millio:. Ib/tr. In asindicated by the process computer. thn event of operativu with a masirium f rhet ion of lima ta ng

2. When the process computer is out of power connity OWtrn) circa t er service, this safety hinit shall Le as, than thu frictio, or ratt.d sumed in be esceeded if the neution power (fvP), the retting nr.nll .

flux cacceds the scram settirip estal>. be modificd an followns lished by Sjecification 2.1.A tind a , rep ]

control roJ scram does r.ot occur, s 6 (.65Wp+ 55) , m.PDJ 1.1/2.1-1 Amendment No. 51

QUAD-CITIES DPR-30 D. Reactor Water Level (Shutdown Condition) where:

FRP = fractiJn of rated Whenever the reactor is in the shut- thermal power down condition with irradiated fuel in the reactor vossel, the water (2511 MWt) level shall not be less than that MPLPD = maximum fraction of corresponding to 12 inches above the li.mit ing power dens-l top of the active fuel

• when it is ity where the J imit-seated in the core. ing power density for each bundle is

• Top of active fuel in defined to be the design linear 300 inches above vessel zero (Sec heat generation rate i Bases 3.2). for that bundle.

The ratio of FRP/MFLPD shall be met equal to 1.0 unless the actu- l al operatir.g value tm less than l 1.0 in which case the actual I operating value will be used.

2. APRM Flux Scrat, Trip Setting (Re.

fueling or Startup and liet Standby Mode)

When the reactor rnode switch is in the Refuel or Startup ifot Standby posi-tion, the APRM scram call be set at less than or equal to !!% of rated neutron flux. i

3. IRM Flux Scram Trip Setting l The IRM flux scram settmg shall be set at less than or equal to 120!!25 of full scale.

4. % hen the reactor mode switch is in the ,

startup or run position, the reactor shall I not be operated m the natural circula.

tion flow mode.

1 B. APRM Rod Illock Setting i The APRM rod block setting shall be as shown in Figure 2.1 1 and shall be; S s (.65Wo+ 43) l 1.1/2.1-2 Amendment No. 51 l

QUAD-CITIES DPR-30 l

l The definitions used above for the APRM scram trip apply. In the event of ope r-ation with a maximum fraction limiting power density (PILPD) greater than the fraction of rated power (FRP), the setting shall be modified as follows:

FRP S 6 (.65Wp + 43) MFLPD The definitions used above for the APPR scram trip apply.

The ratio of TRP to PILPD shall be set equal to 1.0 unless the actual operating value is less than 1.0, in which case the actual operating value will be used.

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

• at normal operating condi-tions.

D. Reactor low water level ECCS initiation shall be 84 inches (+4 inches /-0 inch)  ;

above the top of the active fuel

• at normal operating conditions.

E. Turbine stop valve scram shall be s 10% valve closure from full open.

F. Turbine control valve fast closure scram shall initiate upon actuation of the fast closure sole- I noid valves which trip the turbine control valves.

G. Main steamline isolation valve closure scram  :

shall be s 10% valve closure from full open. 1 H. Main steamline low pressure initiation of main steamline isolation valve closure shall be 2 850 psig.

1 of active fuel is defined to

• be Top 3 60 inches above vessel zero (See Bases 3.2) 1.1/2.1-3 Amendment No. 51 l

i

QUAD-CI' RIES DPR-30 L1 SAWTY LIMIT MS13 The fuel cladding integrity limit is act such that no calculated fuci dt. mage would occur as a a c:;ul* of an abnormal operat ional t rinsacnt. ticcause fuel drmage is not di rec t ly obses vobl e, , a i ttp-back un: Meh is used to establich a safety limit such that the manimum critical power rot to (MCPU) claddirq integr ity saf e ty limit. MCPR > the fuel cladding integrity st4fety limit ri pr ercnt in no elurn than LPeave a coacrvat fuel l margin relative to the conditions required to maintain f uel cledding integrity. I The fuel claddirvj The integrity of this cladding bara t eris one of the phys t eal bart icro which separate rads oret a ve materiots f rom the envi Althoua;h s,mo cot rosion cr use-relatedlacracking migraticn from this source reinted to r: yitsoccur relat ivo f a cedom durirg frcm pe r f ot o t i cins cr crackanq.

the 1)fe of the claddinq, fastion ptc h t is incrementally cumle t tvc and cou tnuovaly rxarurrble. 6 uc t cladding pe r-forations, however, can result f rom t hermal a tt e te n kh tch occur fro:n t eactot operotten significantiv above design pe ectditions r f ors acru a t i on ar d the protectica sy: tem catcty cc t t ing s . Uhilu f aa 0 0n Product W gration frem c!? Hir, at tena t i i s p. t cs measurable as that from urc-reld ed cracking, the oc r : % -

al cladding deterioration. threshold beyond whtch still greater ther*nal stresses envy cause ratrcr thanga onthe trert rmally t outed

.nt-Therefore, the f uct cladding saf ety limj t t r, de f ined w s u ra i g a n to ttc cedi.

tions which t.ould produce onset of transition healing (10PR of 1.0). Tho r.c co ndit i on a repactent a eigntit-cant depart.ur e from inte7tity sa f ety limitthe condit ion int ended by ders t yn for planned oper ation. Thesefore, the f uct el edd ing abnormal operational transient. is cstablished tven that no calculated fuel demaye would occur as a result of an documented in Ec ference 1. Dasis of the values derived for thir,tafety ILmat f or each f uct type as A. Reactor Pressuro > 000 psig and Core riow > IC% of Rated Onset of transition boiling resulto in a decrease in heat t rara f e r f r om the c laddit.y and the re for e elevated claddina temperrture and the possiotlity of cis ddir.g failure. I: owc ve r , t he e x i s. t t r< e o f critical power, or belling transition th not a directly cbservablu paremetet in an operating react-or, Therefore, the margin to boiling trenrition is epiculated frcu pl e nt op e ting parrmetric uch as core power, core flow, feedws ter temperature, and core power ditt: 1but t w . Tr.e margin f or c/ et fuel asserr,bly is char acterited by the critical power ratio (Cit), which t r. tn; ratio of the h.M i c power which would produce onnet of trunnition boiling divided !>y t he actual bmile powt s . 't he sd nimum value of this ratio for any bundic in the core is the minf raum ct it.ical pwer ratio (N rn).

It is assumed that the plant operation is controlled to the norninal protretivt r.rtponts via the instrumonted variables (rigure 2.1-3).

The McFR fuel cladding integrity safety limit has suf ficient con:crvatism to noture that in the tver-l of an rbnormal operataonal trensient ini t iated f rena the normo) ope r ,i t a nq conditivn innre tran 00 X.

of the furi rods an the core are enpoeted to avoid boiling transit ion. The mergin bet w n LC? oi 1.0 (on=et of transition boil 2ng) and the- s.ofety limit, is dc tived f r om u detailed statistical l analyrds considering all of the uncertaintles in hionitoring the core operat1 H ttate, includi q uncertainty in the boiling trcncition correlation (see e.g., kc f crence 1) . Becauur the LN !;r.g trentition correlation is betr4 on a latye quantity of full-scale data, tia re in a very hanh corr )

fidence that operation of a fuel ashembly at t he condit ion of MCPM = the fuc t eledding integrity safety limit would not produce boiling transition. l I

l l

However, if boiling transition were to occur, cladding perforation would not be expected , c3 rdd i ty temperatures would increnne to eppro::imately 1100 r, which in beslew the per f oration t emper nt s t e et the cladding tuternal. This ha* been verified by tests in the Cenars) Flect ric icst U nctor (ct o),

where simil.sr fuci opercted tbove the critical heat flux for a significant petiod of t ime (3C lain -

utes) without citdding perforation.

If reactor pressure r.hould ever execed 1400 psia during norr:n1 power creration (the limit of applicability integrity sofoty of thu limitboiling tranta has been t ion correlation), it would be ncsurred that the fuel cladding violated, ,

i In addition to the boiling trensition Ilmit (MC PR) operation is const rained to a maximum L H cr.s 1*1.5 1 j

kv/ft for 7 x 7 fuci and 13.4kw/f t for ell Dx0 fuel types. This constraint is established by specification 3.5.J.

to p mar in to strain for abnorma.rovide cdequate safety 4operatingtransients. nit 1% p 2atec.,romg.as'ic 198 power conditions. Specification 2.1.A.1 provides for equivalent safety margin for transients initiated from' lower ditions by adjusting the APPM flow-biased scram setgower con-.ing by the ratio of FRP/MFLPD.

Amendment No. 51

l QU Al)-Cl lil;S DPR-30 Speci5 cation 3.5J established the LilGR maximum which cannot be cueeded under steady power operation.

B. Core Thermal Power Limit (Reactor Pressure <800 psia)

At pressures below 800 psia, the core elevation pressure drop (0 power,0 Dow)is greater than 4.56 pst At low powers and nows this pressure dif#erential is maintained in the bypass region of the core. Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low powers and Dows will always be greater than 4.56 psi. Analyses show that with a now of 2S x 10'Ib/hr bundle flow, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi.'Ihus the bundle flow with a 4.56 psi driving head will be greater than 28 x 10'lb/hr. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. At 25% of rated thermal power, the peak powered bundle would have to be operating at 3.86 times the average powered bundle in order to achieve this bundle power.Thus, a core thermal power limit of 25% for teactor pressures below 800 psia is conservative.

C. Power Transient During transient operation the heat flux (thermal power to water) would lag behind the neutron flux due to the inherent heat transler tune cun>iant of the fuel. which is 8 to 9 seconds. Al o. the limiting safety system scram settings are at values which will not allow the reactor to be operated above the Safety limit during normal operation or during other plant operating situations w hich hase been analyzed in detail.

In addition, control rod scrams are suen that for normal operating transients, the neutron flux transient is terminated before a significant increase in surface heat flux occurs, Control rod scram times are checked as required by Specification 4.3.C.

Exceeding a neutron Oux scram setting and a failure of the control rods to reduce Oux to less than the scram setting within 1.5 seconds does not necessarily imply that fuelis damaged; however. for this specification, a safety limit violation will be assumed any time a neutron aux scram setting is exceeded for longer than 1.5 seconds.

If the scram occurs such that the neutron aux dwell time above the limiting safety system setting is less l than 1.7 seconds the safety limit will not be exceeded for normal turbine or generator trips, which are I the most severe normal operating transients expected. These analyses show that even if the bypass system l fails to operate, the design limit of MCPR = the fuel cladding intectrity safety (

limit is not exceeded. Thus , use of a 1.5 second limit provides 1 additional mgrain.The computer proviced'has a sequence annunciation prograr6 which will indicate the seJ ,

scrams occur, such as neutron Dux, pressure, etc. This program also indicates w hen the scram setpoint is cleared. This will provide information on how long a scram condition exists and thus provide some measure of the energy added during a transient. Thus, computer information normally will be available for analyzing scrams; however, if the computer information should not be available for any scram analysis, Specibcation 1.1.C.2 will be relied on to determine if a safety limit has been violated.

During periods when the reactor is shut down, consideration must also be given to water level requirements due to the effect of decay heat. lf reactor water level should drop below the top of the active fuel during this time, the ability to cool the core is reduced. This reduction in core cooling capability could lead to elevated claddinr. temperatures and cladding perforation.The core will be coo!cd sutnciently to prevent cladding melting should the water lesel be reduce 4to two thods the eme height i st.ihhsh-ment of the safety hmit at 12 inches above the top of the fuel provides adequaic margin. 'lhis level will l l be continuously monitored whenever the reciiculation pumps are not operatmg. )

• Top of the active fuel is defined to be 360 inches above vessel (

zero (see Bases 3.2). l l

Amendment No. 51

1. v 2.1. 4 1

1

l l

l QU AI).Cl 1 li'S l)l81( .TU References 1, " Generic Reload fuel Applications," NEDE-24011-P-A*

l l

• Approved revision number at time reload fuel analyses are p'erformed, i I

i i

l 1

Amendment flo. 51 1.1/ 2.1 -6 l

_ _ . . _ _ . _ .. . .. ~ _. . _ , _ _ . . .

Qt ! All-(Tl IF.S '

01 % 30 2.1 LIMITING SAFETY SiSTEM SETTING llASES The abnormal operational transients applicab!c to operation of the units have been analyzed throughout the spectrum of planned operatiag conditions up to the rated thermal power condition of 25 I I MWt. in addition,2511 MW is the licensed masimum ueadpuate power level of the units.This maximum steady state power level will never knowingly be exceeded.

Conservatism is incorporated in the transient analyses in estimating the controlling factors, such as void reactivity coeflicient,. control rod scram worth, scram delay time, peaking factors, and axial power shapes. These factors are selected conservatively with rest'ect to their etreet en the anplicab!c transient results as det; mined b3 the current analysis model. Conservatism incorporated into the transient analysis is documented in Reference 1. Transient analyses are initiated at the conditions given in this Reference.

I ne at> solute value oi ine voto reactivtty coemeient useo in tne anaryns is conservauveiy estimateo to oc anout n.o greater than the nominal maximum value expected to occur during the core lifetime. The scram worth used has been derated to be equivalent to approximately 50?o of the total scram uorth of the control rodt 'Ibe scram delay i

time and rate of rod insertion allowed by the analyses and consarvatively set equal to the longest delay and slowest insertion rate acceptable by technical specifications. The ef!ects of scram worth, scram delay time. and rod insertion rate, all conservatively applied, are of greatest significance in the early portion of the negative reactivity insertion.

The rapid insertion of negative reactivity is assured by the time requirements for 5?o and 20% insertion. By the time the rods are 60'i inserted, approximately 4 dollars of negative reactivity have been insertc<J. which strongly tuins the transient and accomplishes the desired effect. The times for 50?e and 90?o insertion are given to auure proper completion of the expected performance in the earlier portion of the transient, and to cuablish the ultimate fully shut down steady state condition.

This choice of using conservative values of controlling parameters and initiating ttansients at the design power level produces more pessimiuic answers than would result by using expected values of control parameters and analyzing at higher power levels.

Steady.uate operation without forced recirculation will not be permitted except during startup testing.The anal,ssis to support operation at various power and flow relationships has considered operation with either one or two recirculation pumps.

The bases for individual trip settings are discussed in the following paragraphs.

For analyses of the thermal consequences of the transients, the MCPR's stated in Paragraph 3.5.K as the limiting condition of operation bound those which are conserva-tively assumed to ex'ist prior to initiation of the transients.

A, Neutron flus Trip Settings

1. APRM Flux Scram Trip Setting (Run Mode)

The average power range monitoring ( APRM) system, which is calibrated using heat balance data taken during steady state conditions, reads in percent of rated thermal power. Because tission chambers provide the basic input signals, the APRM system responds directly to average neutron ,

flux. During transients, the inuantaneous rate of heat transfer from the fuel (reactor thermal power ) '

is less than the instantaneous neutron flux due to the time constant of the fuel. Therefore, during l abnormal operational transients. the thermal power of the fuel will be less than that indicated by the J neutron flus at the < cram setting. Analyse 3 demonstrate that with a 120Co scram trip setting. none of j the abnormal operational transients analyzed violates the fuel safety limit, and there is a substantial margin from fuel damage. Therefore, the use of flow-referenced scram trip provides even additional ,

rnargin.  ;

! i l

l l Amendment No. 51 l 1

, , - - . , - , . . - , . ,.i

OtiAD-CITI33 DPR-3Cr An increase in the APRM scram trip setting would decrease the margin present before the fuel.ciadding integrity safety limit is reached. The APRM scram trip setting was determined l by an analysis of margine required to provide a reasonable range for maneuvering during )

I o pe r a t ion . Reducing this operating margin would increase the frequency of spurious scrams, which have an adverse effect on reactor safety because of the resulting thermal stresses.

Thus, the APRM scram trip setting was selected because it provides adequate margin for the a fuel cladding integrity sa f ety limit yet allows operating margin that reduces the possibil- l ity of unnecessary scrams.

Tha scram trip setting must be adjusted to ensure that the LHGR transient peak is not ,

increased for any combination of maximum fraction of Ilmiting power density (MELPD) and l reactor core thermal power. The scram setting is adjusted in accordance with the formula l in Specification 2.1. A.1, when the MTLPD is g:ceter than the fraction of rated power (FRP). j l

2. APRM Flux Scram Trip Setting (Refuel or Startup/ Hot Standby Mode)

For operation in the Startup mode 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 l imit , 25% of rated. The margin is adequate to accommodate anticipated maneuvers associated with power plant startuo. Effects of increasing pressure at tero or low void content are i minor, cold water from cources available durir.g otartup la r.ot much colder than tha t already in tho i system, temperature coef ficients are small, and'cbntrol rod patterns are constrained to be uniform by operating procedures backed up by the rod worth minimirer. of all possible sources ,

l of reactivity input, uniform control rod withdrawal is the most probable cause of significant j power rise. Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks, and because several rode must be moved to change power by a signifi- {

s cant percentage of rated power, the rate of power rise is very slow, cenerally, the heat flux '

is in near equilibrium with the fission rate. In an assumed uniform rod withdrewal approach to the scram level, the rate of power rise _ is no moze than 5% of rated power per minute, and the APRM system would be more than adequate 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 switch occurs when reactor pressure is greater than 850 psig.

3. IRM Flux Scram Trip Setting The IRM system consists of eight chambers, four in each of the reactor protection systen logic channels. The IRM is a 5-decade instrument which covers the range of power level between that covered by the SRM and the APRM. The 5 decades are broken down into 10 ranges, each being one-half a decade in aire.

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 divirions f or that range r likewise, if the instrument were on Range 5, the scram would be 120 divisiens 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 the power increase are due to control 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 condition in which the reactor is just suberitical and the IRM system is not yet on '

scale. I Additional conservatiem was taken in this analysis by assuming that the IRM channelC10SOSt to the withdrawn rod is bypassed. The results of this analysis show that the reactor is scrammed and peak power limited to 1% of rated power, thus maintaining MCPR above the fuel cladding integrity safety limit. Based on the above analysis, the IRM provides protection against local control rod withdrawal errors and continuous withdrawal of control rods in sequence and provides backup protection for the APRM, Amendment No. 51 1.1/2.1-8

4 QUAD-CITIES DPR-30 D. 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 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-case MCPR which could occur during steady-state operation is at 100% of rated thermal power because of the APRM rod block trip s e t t i ng . 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 is adjusted downward if the maximum fraction of l imit -

inq power density exceeds the fraction of rated power, thus preserving the APRM rod block 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 instrumentation is density compensated.

D. Reactor Low Low Water Level ECCS Initiation Trip Point The nmergency core cooling subsystems are designed to provide sufficient cooling to the core to dissipate the energy associated withthe loss-of-coolant accident and to limit fuel cladding temperature to well below the cladding melting temper 6ture 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 recutrement 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 components to meet the above criteria was dependent on three previously set parameters: the maximum break sire, the low water level scram setpoint, and the ECCS initiation setpoint.

effective core cooling.To lower the setpoint for initiation of the ECCS could lead to a loss of but To raise the ECCS initiation setpoint would be in a safe direction, i it would reduce the margin established to prevent actuation of the ECCS during normal operation or during normally expected transients.  !

j E. Turbine Stop Valve Scram The turbine stop valve closure scram trip anticipates the pressure, neutren flux, and heat flux increase that could result from rapid closure of the turbine stop valves. With a ceram trip setting of 10% of valve closure from full open, the resultant increase in surface heat flux is limited such that MCPR remains above the MCPR fuel cladding integrity safety limit even during j the worst-case transient that assumes the turbine bypass is closed.

F. Turbine Control valve rast Closure seren The turbine control valve fast closure scram is provided to anticipate the rapid increase in i pressure and neutron flux resulting from fast closure of tha turbine control valves due to a load rejection and subseauent f ailure of the bypass, i.e., it prevents MCPR from becoming less l than the MCPR fuel cladding integrity safety limit for this transient.

1.1/2.1-9 Amendment No. 51

l l

l Q U A D-Cl l li'S DPR-30 G. Reactor Coolant Low Pressure Initiates Main Steam Isolation Yahe Closure The low pressure isolation at 850 psig was provided to give protection against fast reactor depres-surization and the resulting rapid cooldown of the vessel. Advantage was taken of the scram feature which occurs in the Run mode when the main steamline isolation valves are closed to provide for reactor shutdown so that operation at pressures lower than those specified in the thermal hydraulic safety limit does not occur, although operation at a pressure low than 850 psig would not necessarily constitute an unsafe condition.

H. Main Neamline Isolation tu Yahe Closure Scram The low pressure isolation of the main steamlines at 850 psig was provided to give protectien against rapid reactor depressurization and the resulting rapid cooldown of the vessel. Advantage was taken of the scram feature in the Run mode which occurs when the main steamline isolation valves are closed to provide for reactor shutdown so that high power operation at low reactor pressures does not occur, thus providing protection for the fuel cladding integrity safety limit. Operation of the reactor at pressures lower than 850 psig requires that the reactor mode switch be in the Startup position, where protection of the fuel cladding integrity safety limit is provided by the IRM and APRM high neutron f ux scrams.

Thus, the combinateon of main steamline low-pressure isolation and isolation valve closure scram in the Run mode assures the availability of neutron flux scram protection over the entire range of applicability of the fuel cladding integrity safety limit. In addition, the isolation vahe closure scram in the Run mode anticipates the pressure and flux transients which occur during normal or inadvertent isolation valve closure. With the scrams set at 10% valve closure in the Run mode, there is no increase in neutron l flux.

I. Turbine F.HC Control Fluid Low Pressure Scram The turbine EHC control system operates using high-pressure oil. There are several points in this oil system where a lost of oil pressure could result in a fast closure of the turbine control valves. This fast closure of the turbine control valves is not protected by the turbine control valve fast closure scram, since failure of the oil system would not result in the fast closure solenoid vahes being actuated For a turbine control valve fast closure. the core would he protected by the APRM and high-reactor pressure scrams.

However, to provide the same margins as provided for the generator load rejection on fast closure of the turbine control valves, a scram has been added to the reactor protection system which >enses failure of control oil preuure to the turbine control system. This is an anticipatory scram and results in reactor shutdown before any signi6 cant increase in neutron flux occurs. The transient response is very similar to that resulting from the turbine control valve fast closure scram.The scram setpoint of 900 psig is set high enough to provide the necessary anticipatory function and low enough to minimize the number of spurious scrams. Normal operating pressure for this system is 1250 psig. Finally, the control valves will not start until the Guid pressure is 600 psig. Therefore, the scram occurs well before valve closure begirm J. Condenser Low Vacuum Scram Loss of condenser vacuum occurs when the condenser can no longer handle the heat input. Loss of condenser vacuum initiates a closure of the turbine stop valves and turbine bypass valves which eliminates the heat input to the condenser. Closure of the turbine stop and bypass valves causes a pressure transient, neutron Hux rise and nn increase in surface heat flux.To prevent the cladding safety limit from being exceeded if this occurs, a reactor scram occurs on turbine stop valve closure in the Run mode. The '

turbine stop valve closure scram function alone is adequate to prevent the cladding safet, ".mit from being exceeded in the event of a turbine trip transient with bypass closure. i ,

The condenser low vacuum scram n anticipatory to the stop valve closure scram and causes a scram before the stop valves are closed and thus the resulting transient is less severe. Scram occurs in the Run mode at 23 inch lig vacuum stop valve closure occurs at 20 inch Hg vacuum, and bypass closure at 7 inch Hg vacuum.

Amendment No. 51 g

QU AI)-CITif.S DPR-30 References

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

• Approved revision number at time reload analyses are performed 1.1/2.1-11 Amendment No. 51

Figure 2.1-2 has been deleted Amendment No. 51

QUAD-CITIES DPR-30 1,2/2.1 REACTOR COOLANT SYSTEM SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING Applicability: Applicability:

Applies to limits on reactor coolant system Applies to trip settings of the instruments and pressure. devices which are provided to prevent the reactor system safety limits from being exceeded.

Objective: Objective:

To establish a limit below which the integrity of the To define the level of the process variables at which reactor coolant system is not threatened due to an at.tomatic protective action is initiated to prevent overpressure condition. the safety limits from being exceeded.

. SPECIFICATIONS A. The reactor coolant system pressure shall not - A. Reactor coolant high pressure scram shall be exceed 1325 psig at any time when irradiated $1060 psig.

fuel is present in the reactor vessel.

B. Primary system safety valve nominal settings shall be as follows:

I valve at 1115psig'" l 2 valves at 1240 psig 2 valves at 1250 psig 4 valves at 1260 psig

"' Target Rock combination safety /re!ief valve The allowable setpoint error for each valve shall be 1%.

1.2 / 2.2 - 1 Amendment No. 51

QUAD-CllllS DPl{-30 1.2 SAll1 Y 1.htlT D ASI:S The reactor coetant system niter,rity is an important ' arrier in the prnentinn of uncontrol!cd re!casc cf TAon products it is cuential that the interrity of this system be protected by estabbshmp a pressure hmit to be observed for all operating cou.bnons .'nd w heneser there is irraA el f.d in the reactor vene' The prenure safety lunit of 1325 psip a rneasurcJ by the vene! s cam space preuure indicator is equisalent to 1373 PdF at the leurst c!cvanon of the reactor coo! ant system. Tm IU5 psig va:ue n derned from the deupn preuure, of the seauor preuure sene! and ccdant system pip: 1 The respectne ocupn preuures are 1250 pug at 575 ' F and 1115 r if a' I'.0- T . The prenure safety finat was chc,en as ihe lou er ef the preaure trandents permined by t% ;'p".calde (m:n toJes ASMI. Boiler and Pressure Veuel Code Secuen Mi f ar the p<cuure sewel, and USAM M1.1 t.n.e for tN reacter coolant ystem piping The A%1F !!oder and Pn wre Vewel Code ;unuts

' pressure tranuents up to W over daarn preuure (110 ' x 1250 = 1375 pug), and the t 3 %I CoJe pernats pressure tranuents up to 20'.1 os er the deur, pressure (1200 a ll75 = 1110 psic) The safety hmo pressure of 1375 Iniz n referenad to the !*c elesanon of the ertmary coolant u s:em Evaluation performed methodology used to assure that this safety limit ere==ure is not exceeded for any reload is documented in Re:erence 1.

any relond .t 3 documented in Reference 1.

The deurn bau for the reacier pre %ure veuct mal;es evident the sulatannal mart:nr o pr nemm apnst failure at the SJfety preuure hmit of 1375 put 't he vessel has been deugncJ for a pener.tl r umbrane stre:a no preater than 26.7M pu at an tnrernal preuure of 1250 psig, this is a factar of 1.5 belo* t!f peld strenpih of RioO pst at 575

• f. At the prenure hmit of I Mi psig, the general rnembrane stten will only be 29.@9 pu still safeiy t;elow the yield strength.

The relationships of streu levek to yield strength are comparable for the primary system piping and provide a similar nmpm of pro:ecnon at the establoheJ safety pressus hmit.

The nonnal upciatin; r eswre of the sca:toi coolant system is 1030 psig. l'or the turW .e n$ orless of electiicallaad transients, the t .rhme top serrn o fencrator load rejection s: tarn to= ether wah the turn >.e bypass system limits the prenure to arrrouma:ely 1100 ;w (P.cferences 2,3 and4)- In add.non rressure reh<f vahes have been rrov:ird to reduce the probaVny of the s.ifety Des operaw m the event that the turbme bypass shedd fa" '

finally, the safety valves are sired to keep the teactor coo l ant system pres m. l,chiw 1375 psy with r.a credit talen for rebef ulves during the postulat:J full closare of all MSWs without ducu b.,1.s posinon switch) scram. CreJa n t.Len tot the neutron flux scram, howescr.

l 1he indirect fin scram and safety vahe a:tuation, provide adequate marpn l below the pen a: low able veswl prenure of 1375 pwg.

Reactor preuuse is continucosly monnered in the control room during operation on a 1500 psi full scale pressure recorder.

Itefennees

1. "Gencric Reload Fuel Application", N",DC-240ll-P-A*

2. SAR, Section 11.22

3. Quad cities 1 Nuclear Power Station first reload license submittal, Section 6.2.4.2, February 1974. i 1

4. CE Topical Report NEDO-20693, General Electric Boiling Water Reactor No. 1 licensing submittal for Quad Citics Nuclear Power Station Unit 2, December 1974 Approved revision number at timo reload analysco are performed.

1.2/2.2-2 Amendment No. 51 l

AUAD CITIES D PP.- 3 3 e o u.m a t m.p .y r. u,,

e , - - -7. . , , , , . , . , , , - . , , , . , . . . , , . , ,

car:12 c 2 c 12.,. c e t T2..; ancza 1

l In compliance with Sectien III cf tne A M Cede, the safety valves must ce i set to open at no higher than 193f5 of design pressure, and they must limit )

the reactor pressure to no more thar 11375 of design pressure. Both the I high neutron flux scram and safety valve actuation are required to prevent 1 overpressurizing the reactor pressure vessel and thus exceeding the cressure 2 91'ety limit. The pressure scram is available as backup protection to the high flux scram. Analyses are performed as described in the i

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

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

1 1

1.2/2.2-3 Amendment No. 51

QUAD.cIlll:s twR.m.

3.1N.1 IWACTOlt PitOTECTION SYSTEM l.l\lillNG CONDITION 4 F OR OPI'R ATION SUR Yl{ll,1.4 N CI: !!! Ol'lul31tXI S Appht aldhiy : Applicah;hi; Apphe, to the wirumentat:en and am iaied do Apphes ta the sarveClante uf t!.c instrumentation mes w hch initure a rextnr vrm and associatcJ devices w Ns h initia:6 reactrar scrum.

Objectisc: Objecth e:

L awure the cperabilii,s 4 iht rcaom prowtion To specify the type and freque :ry dsarvetilante in ptem.

he app!ied to tre pro:etrien unirumentation.

SPECIFICATIONS A. (be setpoir is, minimum number of trip sys- A. Initrumentation 9 temi shall be functior; ally kms, and niinimum number of instrument tested anJ eahbrated as indicated in Tables channels that must be operchle for each posi- 411 and 412 respectisely Lion of the reauer mode swinh shall he ;n given in Tables 11 1 ihtouch 3.14. The system B. Diily during reacim power operatier the core response times from the opening of the sentor power distributior shallle checkeJ lor maximum toniaci up to and including the opening of the trip actuator centacts shall not exceed 100 fraction of limiting power dens-inilhsecond5- ity (!CLPD) and compared with the I D. If, during operation, the maximum fraction of rated powcr (FRP) fraction of limiting power dens- when operating above 25%, rated thermal power.

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

rated thermal power, either: h is & - ha& I k (A in the unsafe c#ndit an .ad Column 1 or Ta.

1. the APRM scram and rod bles 3.1 1 through 113 unnni t'e met, that block settings chall be tri Psystern must t,e put in the tapp:d condit.on reduced to the values {mmediately. Alict k - R P5 channels that mon-given by the equations stor the same es o!e shall t e funuionally in Specifications 2.1. A.1 teued within A ' .n.rs The trip systcm with the and 2.1.D. failed t hannet raay be untri; p:d far a retied er i time r.o to esceed I hour in snneuet this I testing. At long as the trip system with the failed thaanel sontains at least one operahle channel monitornq that ume variaNe. that trip systeni rnay be placed in the untripped position fer sherl periods of time to allow l

functinnal testini of all RPS instrument chan- i nels as specified by Tatste 4.1 1.The trip <> uem may be in the untripped position (nr 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.at test period for thi,

2. the power distribution "I"I' chall be changed such that the maximum fraction of limiting pcreer density no longer excr*ods the fraction of ratrd powcr.

Amendment No. 51 3.1/4.1-1

t QUAD-CITIES .

DPR-30 gallons. As indicated above, there is sufncient volume in the piping to accommodate the scram without impairment of the scram times or amount ofinsertion of the control rods.This function shuts the reactor down while sufScient solume remains to accommodate the discharged water and precludes the situation in which a scram would be required but not be able to perform its function adequately.

Loss of condenser vacuum occurs when the condenser can no longer handle heat input. Loss of condenser vacuum initiates a closure of the turbine stop valves and turbine bypass valves, which eliminates the heat input to the condenser. Closure of the turbine stop and bypass valves causes a pressure transient, neutron flux rise, and an increase in surface heat flux.To prevent the cladding safety limit from being exceeded if this occurs, a reactor scram occurs on turbine stop valve closure.The turbine stop valve closure scram function alone is adequate to prevent the cladding safety hmit from being exeteded in the event of a turbine trip transient with bypass closure.

The condenser low vacuum scram is a backup to the stop valve closure scram and causes a scram before the stop valves are closed, thus the resulting transient is less sescre. Scram c: curs at 23 inches Hg vacuum. stop valve closure occurs at 20 inches Hg vacuum, and bypass closure at 7 inches Hg vacuum. .

High radiauon levels in the main steamline tunnel above that due to the normal nitrogen and oxygen radioactivity are an indication of leaking fuel. A scrJm is initiated whenever such radiation level exceeds seven times normal background. The purpose of this_ scram is to reduce the source of such radiation to the extent necessary to prevent excessive turbine contamination. Discharge of excessive amounts of radioactivity to the site environs is prevented by the air ejector off gas monitors, which cause an isolation of the main condenser off-gas line provided the limit specified in Specification 3.8 is exceeded.

The main steamline isolation valve closure scram is set to scram when the isolation valves are 10% closed from full open. This scram anticipates the pressure and flux transient which would occur when the valves close. By scramming at this setting. ihe resultant transient is insignificant.

A reactor mode switch is provided which actuates or bypasses the various scram functions appropriate to the ,

particular plant operating status (reference SAR Section 7.7.1.2). Whenever the reactor mode switch is in the (

Refuel or Startup/ Hot Standby position, the turbine condenser low-vacuum scram and main steamline isolation j valve closure scram are bypassed.This bypass has been provided for flexibility during startup and to allow repairs to be made to the turbine condenser. While this bypass is in effect. protection is provided against presure or flux '

increases by the high pressure scram and APRM 15'"o scram, respectively, which are effective in this mode.

If the reactor were brought to a hot standby condition for repairs to the turbine condenser, the main steamline isolation vah es would be closed. No hypothesized single failure or single operator action in this mode of operation can result in an unreviewed radiological release.

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

The IRM system provides protection against excessive pcwer levels and short reactor periods in the startup and intermediate power ranges (reference SAR Sections 7.4.4.2 and 7.4.4.3). A source range monitor (SRM) system is ako provided to supply additional neutron level infor. nation during startup but has no scram functions (reference SAR Section 7.4.3.2), Thus the IRM is required in the Refuel and Startup/ Hot Standby modes. In addition. protection is provided in this range by the APRM !$% scram as discussed in the bases for Specification l 2.1. In the power range, the APRM system provides requirev protection (reference SAR Section 7.4.5.2 ). Thus, the l IR M System is not required in the Run mode. the APRM's r,over only the intermediate and power range; the IRM 's provide adequate coverage in the startup and intermediate range.

The high. reactor pressure, high drywell pressure, reactor low water level, and scram discharge volume high level scrams are required for the Startup/ Hot Standby and Run modes of plant operation.They are therefore required  !

to be operational for these modes of reactor operation.

The turbine condenser low-vacuum t, cram is required only during power operation and must be bypassed to start  !

up the unit.

i d

Amendment No. 51 3,3f a,3 3

e QUAD-CITIES DPR-30 4.1 SURVE!!. LANCE REQUIREMENTS !!ASES

- A. The minimum functional testing frequency used in this specification is based on a reliability analynis using the concepts developed in Reference 1.This concept was specifically adapted to the one out of two taken twice logic of the reactor protection system. The unalysis shows that the sensors are primarily responsible for the reliability of the reactor protection system.This analysis makes use of' unsafe failure' q rate experience at conventional and nuclear power plants in a reliability model for the system. An' unsafe failure' is defmed as one which negates channel operability and which, due to its nature,is revealed only when the channel is functionally tested or attempts to respond to a real signal. Failures such as' blown fuses, ruptured bourdon tubes. faulted amplifiers. faulted cables, etc., which result in ' upscale' or

• downseafe' readings on the reactor instrumentation are ' safe' and will be casily recognized by the operators during operation because they are revealed by an alarm or a > cram.

The channels listed in Tables 4.llt and 4.12 are divided into three groups respecting functional testing.

These are:

1. on-off sensors that provide a scram trip function (Group !): l

2. analog devices coupled with bistable trips that provide a scram function (Group 2); and

3. devices which serve a useful function only during some restricted mode of operation, such as  !

Startup/ Hot Standby. Refuel. or Shutdown, or for which the only practical test is one that can be performed at shutdown (Group 31 ,

I The sensors that make up Group I are specifically selected from among the whole family ofindustrial on off sensors that hase earned an excellent reputation for reliable operation. Actual history on this class of sensors operating in niiclear power plants shows four failures in 472 sensor years, or a failure rate of 0.97 x 10-*/hr. During design, a goal of 0.99999 probability of success (at the 50% confider.cc level) was adopted to auure that a balanced and adequate design is achieved.The probability of success is primarily a function of the sensor failure rate and the test interval. A 3 month test interval was planned for l Group 1 sensors. This is in Leepmg with good operating practice and satisfies the design goal for the j logic configuration utilized in the reactor protection system.

To satisfy the long term objective of maintaining an adequate level of safety throughout the plant l lifetime, a minimum goal of 0.9o99 at the 95% confidence level is proposed. With the one out of two taken twice logic, this requires that each sensor have an availability of 0.993 at the 95% confidence level.

This level of availability may be maintained by adjusting the test interval as a function of the observed failure history (Reference 1).To facilitate the implemtntation of this technique. Figure 4.1 1 is provided l to indicate an appropriate trend in test interval. The procedure is as follows:

1. Like sensors are pooled into one group for the purpose of data acquisition.

2. The factor M is the exposure hours and is equal to the number of sensors in a group, n, times the elapsed time T (M = nT).

3. The accumulated number of unsafe failures is plotted as an ordinate against M as an abscissa on Figure 4.1 1.

4. After a trend is established, the appropriate monthly test interval to satisfy the goal wiil be the test interval to the left of the plotted points.

1 A test interval of I month will be used initially until a trend is established.

Group 2 devices utilize an analog sensor fo!! owed by an amplifier and a histable trip circuit. The sensor o end amplifier are active components, and a failure is almost always accompanied by an alarm and an indication of the source of trouble,in the event of failure, repair or substitution can start immediately.

An 'as is' failure is one that ' sticks' midscale and is not capable of going either up or down in response Amendment No. 51 3.1/ 4.1 -5

. 1 l

l am.

y ,

y _. .__y. . .

j Qll A D-CITIF.S Dl'l(-30 switches, hence calibration is not applicable; ie.. the switch is either on or off.11ated on the above no calibration is required for these instrument thannels.

i B. The MPLPD shall be checked once per day to determine if the APRM scram requires adjustment. This may l  !

normally be done by checking the LPRM readings. TIP traces, or process computer calculations. Only a small number of control rods are moved daily, thus the peaking factors are not expected to change significantly and a daily check of the MFLPD is adequate. l f(cferences

1. 1. M. lacobs.* Reliability of Engineered Safety Features as a Function of Testing Frequency / Nue/ car Safety

• Vol. 9, No. 4. pp. 310 312. July August 1968.

t I

Amendment No. 51

QUAL)-ClllES I)PR-30 4.1 SURVEll.l.ANCE REQUIRI'MENTS B ASI'.S A. The minimum functional testing frequency used in this specincation is based on a reliability analym using the concepts developed in Reference 1.This concept was specifically adapted to the one out of two taken twice logic of the reactor protection system. The analysis shows that the sensors are primarily responsible for the reliability of the reactor protection system.This analysis makes use of

• unsafe failure' rate experience at conventional and nuclear power plants in a reliability model for the spiem. An ' unsafe failure' is denned as one w hich negates channel operability and which. due to its nature, is revealed only when the channel is functicnally tested or attempts to respond to a real signal. Failures such as blown fuses, ruptured bourdon tubes, faulted amphners. faulted cables, etc., which result in

• upscale

• or

'downscale' readmps en the reactor instrumentation are ' safe' and will be easily recognized by the operators during operation because they are revealed by an alarm or a scram.

The channels listed in Tahie 4.1 1 and 4.12 are divided into three groups respecting functional testing.

These are:

1. on ofTsensors th.n provide a scram trip function (Group I);

2. analog devices coupled with bistab!c trips that provide a scram function (Group 2); and

3. devices which serve a useful function only during some restricted mode of operation, such as Startup/ Hot Standby. Refuel. or Shutdown or for which the only practical test is one that can be performed at shutdown (Group 3 t The sensors that make up Group 1 are specifically selected from among the whole family ofindustrial on-off sensors that have earned an excellent reputation for reijable operation. Actual history on this class of sensors operating in nuclear power plants shows four failures in 472 sensor years, or a failure rate of 0.97 x 10 */hr. During design, a goal of 0 99999 probability of success (at the 507e confidence level) was adopted to auute that a balanced and adequate design is achieved. The probability of success is primarily l a function of the sensor f,ulure rate and the test interval. A 3 month test interval was planned for i Group 1 sensors. This is in keeping with good operating practice and satisfies the design goal for the I logic configuration utilized in the reactor protection system. j To satisfy the long term objective of maintaining an adequate level of safety throughout the plant lifetime, a minimum goal of 0.9999 at the 957v confidence level is proposed. With the one out of two I taken twice logic, this requires that exh sensor have an availability of 0.993 at the 95"o confidence 1:sel.

This level of availability may be maintained by adjusting the test interval as a function of the observed failure history ( Reference 1). To facilitate the implementation of this technique, Figure 4.1-1 is provided l to indicate an appropriate trend in test interval. The procedure is as follows:

1. Like sensors are pooled into one group for the purpose of data acquisition.

2. The factor M is the exposure hours and is equal to the number of sensors in a group, n, times the elapsed time T (M = nT).

3. The accumulated number of unsafe failures is plotted as an ordinate against M as an abscissa on l Figure 4.1 1.

4. After a trend is established. the appropriate monthly test interval to satisfy the goal will be the test l interval to the left of the plotted poinit

5. A test interval of I month will be used initially until a trend is established.

Group 2 devices utilize an analog sen'.or followed by an amplifier and a histable trip circuit. The sensor and amplifier are active components, and a failure is almost always accompanied by an alarm and an indication of the source of trouble,in the event of failure, repair or substitution can start immediately.

An 'as is' failure is one that ' sticks' midscale and is not capable of going either up or down in response Arnendment No. 51 3.1/41-5 i

QUAD-CITIES DPR-30 3.21.1%!! TING CONDillO% i 011 OPi IMTION IMisi.S in addnien b reactor prn:wtion instrum(nunon w hich initiJtes J reattor wram proicuis e in>trun.cntaiam t.n been provid:J which inoutes acuen in net @ne the tomequences of accidena w hah are beyond the oferdlof \

Abty to wntr4 or teunmatn operator errms t eiere they result in sermt.s conseqtaenu h set of spms un -

provides the hminng conditions of operation for the primary syttem isolatun function. imuation ei the emerycosy core cochng system. wntrol rnJ block and sundby rat treatmere systems.1he objectnts of the spetiku am are

(!) to assure the effectnenew of the protetthe imtrutnenutica when required by preserving m e.ipAhty to to! crate a sing!c f.dlere of any wmponent of such systems o en during periods w ben porhons of sa h syueno are out of service for maiatenarae. and (2) to prescr.he the inp settings required to auure adequate lmformance.

% cen necesury. one thannel may be nude inoperable far hueriniervals to conduct required funenonal tesn and cahbratiom Some of the setimp on the instrun enuuan that initiates or controh core and containment talmg h.ne tclerantes expheitly suttJ where the high and low values are both ersucal nd may h.ne a sub untial etist on safety, it shnulJ be noted tnt the setpnints of other mstrun entation where only the hyh er low end of the senig has a 6teet bearmt on ufety are thosen at a loef away from the normal oper.oin; ranye to prewnt inadscrtent actuation of the ufety sys:ern invohed and esposure to abnormal situanent botauon vahes are insu!!cd in those hnes that renetfd'e 'bf Primary ontainment and r,unt Le iwlatcJ durmy a lossef coolarn amdem so tr at tbc r.ubation dme knnts are not citeeced dating an amdent wndinon. Actualmn of thesc salves n inman"t hy the profeethe imu ument unm w hn h ... .s ih wndshorn for w hu h nol.nion n required (thn Mrumentauon is sh.m n in Ianle 3.2 11. Sud mstrumenunon musi he anulaNe w hcnes er prnnary containment integrity a ryired The objectise n to nolate the primary tenuinment so that the guidel.nes of 10 Cl R 100 are not encerJed Jurmy an asciJeni.

The instrumenuuan whwh iniuates prirnary syuem notation n ennnetied in a dual but arrangement. T hu ihe dacuaion gnen in the bases for Speedication 3.1 n appheaNe here.

The low-reactor water level instrumentation is set to trip at > 8 inches on the icvel instrument (top of active fuel is defined to be 360 inches above vessel rero) and after allowing for the full power pressure drop across the steam dryer the low level trip is at 504 inches above vessel vero or 144 inches above the top of active fuel. Retrofit Ox8 fuel has an active fuel 1cngth 1.24 inches longer than earlier fuel desic,ns. However, prcsont trip setpointo were used in the LOCl> analyses (NEDO-24146A, April 1979) .

This trip initiates closure of Group 2 and 3 primary containment isolation valves but does not trip the recirculation pumps (refer-enco SAR Section 7.7.2). For a trip setting of 504 inchos above vessel r.cro (144 inches above top of active fuel) and a 60-necond valvo closure time, the valvos will be closed before perforation of the cladding occurs even for the maximum breaks the setting is therefore adequate.

The low low reactor level instrumentation is set to trip when reac-tor water level is 444 inches above vessel zero (with top of activo fuel defined as 360 inches above vessel zero, ~59 inches is 84 inches c.bave the top of active fuel) .Thn trip initutes slowre or Group i primary ennuinment imlanon

)

s h es (reference SAR hetuur. 7.7.2.2) and abo atuvates the ! CC subtystems. starh the emer-lency stiesel l 1

renerator. .n.J inrs the retirculation pumps.1hu trip setting lesel w as chcsen to be high enough to present spurious operanen but low enough to ininate FCCS operauan and prirnary sptern isolation so that no rusiain; oithe tuel claddmg w att ou cr and to that pmi..ctsJent wohny can be accomp!nhed and the puiJebnes of 10 C F R IDO wdl rmt be eteceded for the comp!cte cirt'umferential break of a 28 inch reetreulation hne and with the trip setting gnen above. I O h inin. tien and [ rimary system notation are initiated and in tune in meet the ahme t ritena

. Ihc instrumenution aho cmers the full spectnnn of breaks 1 and meets the ahose cuieria 1

The high dr>well pressure iratrumentation is a hatkup to the waier level inurumentation and, in addiiion to ininatir.g ECCS. it causes nolation of Group 2 nolation vahes For the breaks docuv.ed above. this irntrumenta-tien willinitiate ECCS operation at about the same t me as the low low w ater levelirmrumentation; thus the resuhs given above are opplicaNe here aho Group 2 holaunn s ahes include the drywell vent. purge. and surnp holanon nahet fligh dr)scli preuure acuvates only these vahes because high drywell pressure could octur as the result

1. 4 non safety rcined causes such as not purging the drywell air during uariup. Total sprem holation i noi Jewahle for these sonJmont and only the vahes in Group 2 are required to close. The low low water leul tenumentanon inniates prowtuon for the luu spettnun oflowof coolant accidents and caum a trip ur Group 1 punary spiem no!ation sahet j L

Amendnent No. 51 ,

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QUAD-CITIES DPR-30 k omo tubes are provided in the main steamlines as a means of measuring steam now and ako limiting the loss o' man inventory from the seuci during a steamline break accident. In addition to momioring steam now, imirumentation is provided which causc> a trip of Group i isolation vahet The primary function of the imo unenianon h to detect a break in the main steamline. thus only Group 1 valves are dosed. For the worstwe act: dent, main ueamline bre1L nutude the drywell. this trip setting of 120% of rated steam now,in conjunction woh the now limuers and main ueamline vahe closure. limits the mass inventory low such that fuel is not unwvered, fuel temperatures remain less than 1500' F, and release of radioactivity to the environs is wcl! below to CFR 100 guidelines (reference SAR Sections 14.2.3.9 and 14.2.3.10).

Temperature monitoring instrumentation is provided in the main steamline tunnel to detect leaks in this area.

Trips are provided on this instrumentation and when escecded cause closure of Group i isolation valves. Its setting of 200' F h low enough to detect leaks of the order of 5 to 10 gpm; thus it is capable of covering the entire spectrum of breaks. For large breaks, it is a hackup to hiphoteam now instrumentation discuued above. and for stuall breaks with the resuhing srnall release of radioactivity, gives isolation before the guidelina of to ('l R 100 are esteeded.

Ihgh radiation monitors in the main steamline tunnel have been provided to detect grou fuel failure This instrumentation causes closure of Group i valves, the only valves required to dose for inis aicident. With the estahinhed setting of 7 times normal background and main steamline isolation vah e closure, 6uion product release is linuted so that 10 Cl:R 100 guidelines are not exceeded for thk ate; dent (reference SAR Secticn 12.2.1.7 ).

Pressure instrumentation is provided which trips when main steamline preuurc Jrops below Sfu psig. A trip af thh instrumentation results in closure of Group 1 isolation valves. In the Reit.el and 'itartupillot hiandM mode-inis trip function is hvpaaed This function k provided primarily to provide protection acainst a preuure repul uor malfunction which would cause the control and/or hypan valve to open. With the trip set at 950 psig, imentory loss is limited so that fuelis not uncovered and peak claddine temperatures are much less than 1500' F, thus. there are no 6ssion products available for release other than those in the reactor water (reference SA3 Section  ;

i 1.2.3 )

The RCIC and the llPCI hi hE now and temperature instrumentation are provided to detect a break in their respecthe piping Tripping of this instrumentation results in actuation of the RCIC or ofIIPCI isolation valves Tripping logic for this function is the same as that for the m. ia steamline isolation vah es, thus all sensen ars required to be operable ur m a tripped condition in meet the sincle tailure cruena lhe nip senmp of M F and 3no~c of design now and valve closure time are such that wre uncosery i prevented and 6ssion produst release is within hmits.

The instrumentation which initiates ECCS action k arranged in a one out of two taken twice logic circuit. Unlike the reactor scram circuits, however, there is one trip system associated with each function rather than the two trip 1 systems in the reactor protection system.The single failure criteria are met by virtue of the fact that redundant ente wohng functions are provided. e p., sprays and automatie blowdown and high pressure coolant injection. ihe speci6 cation requires that if a trip system becomes inoperable, the system u hich it activates is dedared noperable.

For example,if the trip system for core spray A becomes inoperable. core spiav A is dedar d inoperable and the

.*ut ofaervice speci6catmns of Speci6 cation 3.5 govern. This specincation pre >enes the cifeuhenew of the system with wspect to the single fadute crueria esen during periods when mainienance or testinc is being performed.

The control rod block functions are provided to prevent exceuive control rod wid rawal(o that MCI'R Joes not go below' the MCPR Fuel Claddipo Integrity Safety Limit.

The trip logic for this function is one out of n; e g., any trip on one of the sis APRM *s. eight IRM N l four SRM's will result in a rod block The minimum instrument channel requirements anure subent l i

instrumentation to assure that the single failure erderia are met.The minimum inurument channel requirements for the RBM may he reduced by one for a short period of time to allow for maintenante, testing, or calibration.

Thh time period h only-3% of the operating time in a month and does not signi6cantly increase the risk of preventing an inadvertent control rod withdrawal.

1 Amendrnent No. 51 3.2 n.2 d

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QUAD-CITIES DPR-30 so that none of the activity released during the refueling accident leaves the reactor building via the normal ventilation stack but that all the activity is processed by the standby gas treatment system.

The instrumentation which is provided to monitor the postaccident condition is listed in Table 3.2 i. The instrumentation listed and the limiting conditions for operation on these systems ensure adequate monitoring of the containment following a loss-of-coolant accident. Information from this instrumentation will provide the operator with a detailed knowledge of the conditions resulting from the accident: based on this information he can make logical decisions regarding postaccident recovery, The specifications allow for postaccident instrumentation to be out of service for a period of 7 days. This period is based on the fact that several diserse instruments are available for guiding the operator should an accident occur, on the low probab.lity of an instrument being out of service and an accident occurring in the 7-day period, and cn engineering judgment.

1 The normal supply of air for the control room ventilation system comes from outside the service building. In the  !

event of an accident, this source of air may be required to be shut down to prevent high doses of radiation in the control room. Rather than provide this isolation function on a radiation monitor installed in the intake air duct, )

signals which indicate an accident 'i.e., high drywell pressure, low water level, main steamline high flow, or high )

radiation in the reactor building ventilation duct, will cause isolation of the intake air to the control room. The above trip signals result in immediate isolation of the control room ventilation system and thus minimize any radiation dose.

P Amendment tio. 51

I 1

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l OUlsD-CITIES i D rit-30 i

The ?.PRM rod block function ic flow bincod and provents a r:ignifiennt reduc t ion in ".C PR , cepecially during operation nt reduct d flow. Th" APPd providen gross core pro 6ction, i.e., limitn the grecc of control roclo in the normal w Rhdrawal sequence.

In the refuel and startup/ hot etnndby niodes, the APRM rod block function in r:et at 12% of rated power. Thin control rod block provides the como type of protection in the Raftici and Startup/ Hot Standby raoden as the APRM flow-birmed rex 3 bloc 1: dem o in the kun rnodo, i.e., preventn control rod withdrawal beiore a nernm in reached.

The RDM rod block function prov.irlen local protection of the core, i.e.,

the prevention of trancition boiling in a local retrion of the core for a single rod withdrcwnl error f rom a J imiting control rod pattern. The '

trip point in flow biased. Tho storst-care singic control rod withdrawal error is analyred for erach reload to assure that, with the npocific trip l rettings, rod withdrawal is blocked before the MCPR reachen tho' fuel i

cladding intectrity enfety limit. I Below 30X power, the worst-caco withdrawal of a single control rod with-out rod block action will not violate the fuel cladding inteority t nnfety limit. Thun the twM rod block function 10 not recruired below thin I power 1cvel.

The IPd block function providec local as well as groso core protection.  !

The scaling arrangenent is such that the trip setting is less than a factor of 10 above the indicated level. Analysin of the worrit-caco accident resultn in rod block action before MCPR approaches the MCPR fuel cladding integrity safety limit.

A dov.tacale inh:km on an APlt \f or IKN! n im indican,n the instrumcnt has f.n:cd or is r.uc sendtive enough.

In either case the isnitument u di nnt respand to changes in c omrol rod motion, and the contro! rod twtion is thus prevented The doa nscale inps are et ai 3/125 of foil tale, .

The SRN! rod blad unh F 103 0PSand the deresior nei fu!!y inserted assures that the SR\t's are rot withdrawn frorn thc core poor ta comrnerning rod withdrawal foi startup 1he strain dacharpe volume hyh uen t lesel rod tlod prow annuncianon for opriator action lhe alum se: point has been scksitd to provide cdnjuste t:mt to sliow deterrmnahon of the cause oflevel increase and coricctive action prior to automane scram initianon for cRecth e ernertency core conhnr for srnal! ripe breb. the llPCI systern most function. since reactor pecuure does not deerece rapWy ennuth to aMow enher core spig or ! PCI to opet:.e in tune 11 t autoc rine pressure rehef funcuon n pmuded as a had cr to the HPCIin tL ornt the llPCI does not opera'e ll.e airangement of the tripping contaca n mb .n to pmude this function w Len ncecuary and mmimin spurmus operanon lhe top senir'[s f turn in the Spt*cdicahan nr sideqmite to vuure the ahns e Criteria are met (refetcDr f h4 N \/t lion 6 2 6 3 )

lhe spetuieation preserves the clinineness of the sptem during periods of mainten,n.sc. toin? or (abbration and aho nnnunges the tal of inadsenent operatun. i c., only nne instrument channel out of wavice.

Two au ejutor ou ras monnon .or proudsd and, when their trip pomt is reached, cause au nal.unm of the air ejector nn ps hne lwhmon n ' nonated w ben both nmunnen% ec n h their high trip pinnt or one he an upscale trip cd the other a downuale inp 1 hoe is a I$ nnnuic dctiy beloie the air ejcetor ofty.n ini uion vahe n(lmed.

Thn Jc ay is accounted hir by ihe in mmute hoIJop inne of the oCqat before it n rcle.ned ta ihe chimney Hoth anuroments are requised for top, but the unntanenh are so designed that any snurument f nhire rives a dowrneate top lhe inp scionp of the untrumcnn arc set so that the chimney rele ne raic knui pven in Speuhcation 3 5 A 2 n r6ot euceded four radutbn monitors are prouded m the f eatior building sentilation ducts whkh inuiate isolatmn of the reactor buddmr and operatmo of the standby f.n treatment sptem, lhe monitors are Im atcJ in the reactor buddny senidaiion dm i li.e inploric n a one out of ino hir cash set, and each set can inniate a inp andependent of the other vs Any up+de inp s dl sause the demed aciaan 1 rip senings of 2 rnR/hr for monnon in the ventilaiion dot t ,nc based og n annutmr om nul n enhlahon ho! umn and standby ps tic.ument system operanon 50 elut tf e s vnuta m st.nl Nic.m taic knni rn en in Sp. con ahon 3 f.A 3 is ont euceded, Two radi omo nomuors

\

ore prosidcJ on the reinene floor w hnh uniun i.otation of the reacior budding and operatnm of the stanJhy '

pn t caimeni syusms lbr inp los n one out of two lop settings of 100 rnR/In for the anoncon on the refuehnt fWr .uc b.ncJ upon nuiunny noun d ventdanon nolainm and uandby pn trcanntni ytrin operanon l

Amendment No. 51 3 De e

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QUAD-CITIES DPR-30 l

l TABLE 3.21 INSTRUMENTATION THAT INiilATES PRIMARY CCNTAINMENilSOLATION FUNCTIONS maimum uenber of Operath or Tripped Instrument Channelilli lastruments Trip' Level $stting A ct'on2) 4 Reactor low water * >ld4incres above top of A active fue*

4 Reactor low low water d'.4nches above top cf A actr.e fuel

• 4 Hgh drywell pressure

• 52pst* A 16 H(h flow man steamline* s120% of rated steam flew 6 16 High temperature man s200' F steamline tunnel 4 High radiaten man s7 x ncrmal rated power 8 steamine tunnel'H background 4 Low man steam pressure'" 2850 pst B 4 Hqrh Row RCIC steamine 5300% cf rated steam flow C 16 RCIC turbne area h!gh s200

• F C temperature A Hqrh flow HPCI steamline $300% of rated steam flow D 16 HPCI area high temperature s200' F D Mirtes

1. Whenever prunary contamment etetrity is resumed. there shall be two operabis or tr$ ped systems for each functen, except lot bw pressure mam steamime whch only need be available e the Run positen.

2. Acten: If the last column cannot be met for one of the trip systems that tre system shall be tr$ ped.

ll the Erst column cannot be met for both trip systems, the appropriate actees Isted beber shelf be taken.

A, bitute an orderly snutdown and have the reactor m Coid Shutdown conditen a 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

8, initute an orderly lead redveten and have reactor m Hot $tandby withe 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

C. Cbse makten valves a RCIC system.

Q. Cbse aclaten wahres a HPCI subsystem.

3. Need not be operable when primary contammerit stegftty is not requead

4. The notaten tirp signal u bypassed when the mode switch is m Re'uel or $tartup/ Hot $hutdown.

$. This estrumentaten also isolates the control reorn ventilaten system.

6. The sqnal aho automatcally cbses the mechancal vacuum pump discharge idle isolaten wahres.

• Top of active fuel is defined as 360" above vessel zero for all water levels used in the LOCA analysis (See Bases 3.2).

3 m 2-Il Amendment No. 51 l

QUAD-CITIES DPR-30 TABLE 3.2 2 INSTRllMENTAll0N THAT INiilATES OR CONTROLS THE CORE AND CONTAINMENT COOLING SYST[MS Minimum Manaer of Oper:We er fr(pped Instrument Channets* Trip Function Trip Level Setting Remerns 4 Reactor low low 2L4 inches ( 4 8 inches /.0 inch) 1. In conjuncter, with low reactor pressure water level aoove top of actrve fuel

• hitiates core scray and LPCI.

2. In conjuncten with h:g%drywell pressure ,

120 second tee delay and low. pressure core coolirg eterlock hitiates auto b!cwdown.

3. Initiates HPCI and RCIC.

4, initiates starteg of diesel generators.

4'" Hth-drywell s2 psig 1. Initiates core spray, LPCI, HPCI. and pressure'21,m 3GTS,

2. In conjunction with low low water level, 120 secend tee delay, and low pressure core eco;rg hierlock hitiates auto blowdown.

3. Initiates starting of diesel generators.

4. Initiates isolaten of control room ventilation. i Reactor bw 300 ps:gsps350 psig 1, Permissive for openog core spray and LPCI 2

pressure admission vanes.

2. In cocjuncton with low low reactor water l level hitiates core spray and LPCI.

Contahment spray Prevents inadvertent operaten of containtnent hterlock Spray during accident conditons.

2* 2/3 core height 22/3 core height 45 contanment 0.5 psigsps1.5 psig )

1 high pressure l

2 Tener auto s120 seconds in conjunction with low low reactor water i blowdown level, high.drywell pressure, and low. pressure core coolict interlock hitiates auto blow.

down.

4 Low. pressure core 75 psigsps100 psig Defers APR actuation pending confirmation of coolog pump dis. bw. pressure core cooling system operaten.

charge pressure 2 lJndervoltage en N/A 1. Initiates starthg of d'sel e generators.

emergency buses 2. Pe missive for startrg ECCS pnps.

3. Removes nonessential loads from buses.

• Top of active fuel is defined as 360" above vessel zero for all water levels used in the LOCA analyses (See Bases 3.2).

3.2/4.2-12 Amendment No. 51

QUAD-CITIES DPR-30 TABLE 3.M INSTRUMENTATICN THAT INiflATE3 R00 BLOCK Manwm lie'er of Operar,te or Tripped lastrument Channets per Trip $ntem D instrument trto tevet iettet 2 APRM upscale (now bias /4 FRP 2 APRM upsca'e (Refuel and Startup/Het 6[0.650W D

$12/125 full sca:e

+ 43] MFLPD Standby mede) 2 APRM downscaV" 23/125 futi scsk 1 Rod bioch moniter upscale (f.ow biasyn s0 650W + 42m 1 Rod bbck monitor downscaWH 23/125 full scale 3 IRM downscale *

• 23/125 full scale 3 IRM upscak'n

$108/125 full seat 2(U SRM detector not e Startup positon* 22 feet below core center.

Ine 3 (RM detector not in Startup positon* 22 feet bebw core center.

fee 2*

• SPM upscale

$105 counts /see 2* SRM downstaWS 2102 counts /sec 1 Hah water level e scram dischstge volume s25 galbns Notes,

1. For the Startup/ Hot Standby end Hun peritions of the reactor mode cclecter switch, tion except there thuchall SRMbe rodtwo opmabic or tripped trip systema for coca fune-bioche. IRM upscale and IRM downscale riced riot be opuruble in the !<un paition ApuM downweale, APRM upscale (fuu Siwd),

pnd RhM doimealc need not be operable in the Startup/ Hot Standay mo irn l Ttw hiW upscale hved not be operable u t, loss than 30% t'ated thacuni power.

Orie ebenriel may be byparsed above 3M rated thermal power provided t ha t; a limitinc control rod tuttetti does not exist. For sys temo wi t.h nre than one charinal per trip syst.om, if the firs t column cannot be met for one of tbc two trip sys tema , this cond! tion mal/ cxist for up to 7 days provided that d:tring tha t titao the oporonic system is functionally tested im-media tely and dhily ti oreaf ter; If thic condition lastu longer than 7 days the cyskm chu11 be tripped. If thu firs t column cannot be met for both trip r+ ten, the pystems chall be tripped.

2. Wd in stie percent of drive flow tequired to produce a rated core ficw of 90 millien Ib/hr. Trip Icvel setting is in percent of rated power (2511 M4t).

3 utM essex4ie mey tw bysmed when it a on its beest ante 4 the functon a bypassee when the count rate 4 2100 CPS.

5 one of the but MM anputs may be bypessed 6 Tbs MM knctan may be bypassed a the highof IRM ranges tranges 8. 9, and 10) nhen the IRM vpscale rod bbet is ocerabie L bt segewed to be operatie ende perbreeg low power physics tests al atmospheric pressere dereg or etter re4enet at power levels not to escoed 5 M4t 8,

IMs l#M knCten occurs When the '#8ctor mode SettCA es m the Refuel or startup/ Hot $tendby ponden

'9 The trip a bypassed when the SPM rs Nily mserted Amendment No. 51 3.2/4.2 14

_ _ . . _ _ _ _ _ _ _ _ _ . _ . . _ . . m . _ . .

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• QUAD-CITIES DPR-30 TABLI 3.24 P051 ACCIDENT MON 110 RING INSTRUMENTAfl0N RIOUIR[MENTS$

lastrument Mmmum alveter Reedoet of Opersbee tocation Nomter Channelsmm Parameter Unit 2 Praded Range 1 Reactor pressure 902 5 1 01500 psig 2 0-1200 ps.g i Avector wster level 902 3 2 100 inches + 200 inches (0 inches is top of fuel)

• 1 Torus water temperature 902 21 2 0 200* F 1 Torus as temperature 902 21 2 0-600* F Torus weter level, 902 3 1 25 inches - e 25 inches '

2 tti 6ndicator Torus weter level, 1 18 inch range sight glass 1 Torus pressure 902 3 1 5 inches Ng to 5 psig 1 Orywell pressure 902 3 1 5 inches Hg io 5 psig 0 to 75 psig 2 Oryweil tempersture 902 21 6 0 000* F 2 Neutron monitoring 902 5 4 0.1 10' CPS 2 IU Torus to drywell 2 43 psid dif f orentiel pressure

. I Notes 1

1. Instrument channels required during power operation to monitor posteccident cond.tions.

2. Provisions are mede for local semphng and monitoring of drywell atmosphere.

• Top of active fuel is defined to be 360 inches above vessel zero (See Bases 3.2).  !

l i

l Amendment flo. 51.

QUAD-CmES DPR-30

3. The control rod drive housing support

3. The correctness of the control rod system shall be in place during reactor power operation and when the reactor -

withdrawal sequence input to the RWM computer shall be veri 6ed after coolant system is pressurized above loading the sequence.

atmospheri: pressure with fuel in the reactor vessel, unless all control rods Prior to the start of control rod with, are fully inserted and Specification drawal towards criticality, the capabil.

3.3.A.I is met. ity of the rod worth minimiter to a.

properly fulfill its function shall be Control rod withdrawal sequences verified by the fo!!owing checks:

shall be established so that max.

imum reactivity that could be a. The RWM computer online diag. '

l added by dropout of any incre. g;; g , 33g; rnent of any one control blade Performed.

would be such that the rod drop accident b. Proper annunciation of the selec-design limit of 280 cal /am. _is _not exceeded. tien error of one out of sequence

b. Whenever the reactor is in the e ntrol rod shall be verified.

Startup/ Hot Standby or Run c. The rod block function of the rnode below 20% rated thermal l RWM shall be verified by with.

power, the rod worth minimizer drawing the first rod as an out.

shall be operable. A second opera. of sequence control rod no more tot or qualified technical person than to the Nock point.

may.be used as a substitute for an inoperable rod worth minimizer which fails after withdrawel of at least 12 control rods to the fully withdrawn position. The rod worth minimizer may also be bypassed for low power physics testing to demonstrate the shut.

down margin requirements of Specification 3.3.A if a nuclear engineer is present and verifies the step-by. step rod movements of the test procedure.

4. Control rods shall not be withdrawn 4. Prior to control rod withdrawal for for startup or refueling unless at least startup or during refueling, verify that two source range channels have an at least two source range channels observed count rate equal to or greater have an observed count rate of at least than three counts per second and these three counts per second.

SRM's are fully inserted.

5. During operation with limiting con. .

5. When a limiting control rod pattern l

trol rod patterns, as determined by the exists, an instrument functional test of nuclear engineer, either:

the RBM shall be performed prior to

a. both REM channels shall be withdrawal of the designated rod (s) and daily thereafter. l operabic, '

b. control rod withdrawal shall be blocked; or Amendment No. 51 3,3f 4,3 3 i

,. . _ . . -- . ~- - - - . - . .- . -.

QUAD-CITIES I)PR-30 *

c. the operating power level shall be limited so that the MCPR will re-main above the MCPR fuel cladding l l

integrity safety limit assuming a sin- I gle error that results in complete withdrawal or any single operable control rod.

C. Scram insertion Times C. Scram insertion Timer

1. The average scram insertion time, ha- 1. After refueling outage and prior to sed on the deenergitation of the scram operation above 30"r power with re. <

pilot valve solenoids at time rero.of all _ actor pressure above 800 psig, all con-  !

operable control rods in the reactor trol rods shall be subject to scram time j power operation condition shall be no measurements from the fully with.

greater than: drawn position.The scram times shall be measured without reliance on the Average Scrorn control rod drive pumps.

% inserted From Insertion Fully Withdrawn Times (sec) 5 0.375 20 0.900 50 2.00 90 3.50 The average of the sc. ram insertion times for the three fastest control rods of all groups of four control rods in a two by two array shall be no greater than-

)

% Inserted from Average Scram l Fully Withdrawn Times (sec) 5 0.39N 20 0.954 50 2.12 90 3.80 I l

2. The masimum scram insertion time 2. Following a controlled shutdown of for 90% insertion orany operable con- the reactor, but not more frequently trol rods shall not exceed 7 seconds.- than 16 weeks not less frequently than

. 32 weck intervals $0% of the control

3. If Spectneation 3.3.C.! cannot be met.

rod drives in each quadrant of the ,

the reactor shall not be made super-reactor core shall be measured for the I critical: if operating, the reactor shall scram times specified in Specification be shut down immediately upon deter 3.3.C. All control rod drives shall have mination that average scram time is experienced scram test measurements O N#I'*'

cach year. Whenever all of the control

4. If Specification 3.3.C.2 cannot he met, rod drive scram times have been mea.

the de6cient control rod shall he con. sured, an evaluation shall be made to 3'3'd'3 ~ d Amendment No. 51

, s r

l

)

1 J

QUAD CITIES DPR-30 B. Control Rod Withdrawal

1. Control rod dropout accidents as discussed in Ecfcrence 1 con lead l to ci6nificant core damacc. If coupling integrity is maint.uined, the possibility of a rod dropout accident is climinat.ed. The over-travel position feature provides a positive check, as only uncoupled drives may reach this position.

~

Neutron instrumentation response to rod movement provides a verification that the red is following its drive. Absence of such response to drive movement would indicate an uncoupled condition.

2. The control rod housing support restricts the outward movement of a control rod to less than 3 inches in the entremely remote event of a housing failure.The amount of reactivity which could be added by this smali amount of rod withdrawal, which is less than a normal single withdrawal increment. will not conitibute to any damage to the primary coolant system The design basis is riven in Section 6.6.l. and the design evaluation is given in Section 6.6.3 of the SAR. This support is nnt required if the reactor coolant system is a t atmospheric pressure. since there would then be no driving force to rapidly eject a drive housing. Additionally, the support is not required if all control rods are fully inserted or if an adequate shutdown margin with one control rod withdrawn has been I demonstrated. since the reactor would remain suberitical even in the event of complete ejection of the strontest control rod.

3 control rod withdrawal and insertion sequences are established to assure that the rnaximum insequence individual control rod or control rod scr,monts which are withdrawn could not be worth enour,h to cause the rod drop accident design limit of 280 cal /cm to be exceeded if l they were to drop out, of the core in the manner defined for the rod drop accident. These acquences are developed prior t,o initial opor-ation of the unit following any refuelin6 outage and the requirement that an operator follow thene ncquences is supervir.ed by the HWM or a occond quu))fied otation cmploycu. These acquences are developed g to limit reactivity worths of control rods and , [

togeiher with the integral rod vehicity hiniters and the action of the control rod drive system.

limits potential reactivity insertion such that the resuhs of a control rod drop accident will not esteed a maximurn fuel energy content of 280 cal /gm lhe peak fuel enthalpy of 280 cal /gm is below the enert,y content at which rapid fuel disperul and primary system darnage hac been found to occur based on experarnental data as is discussed in Reference 2 , j The analysis of the enntrol rod drop accident was originally prese'ited in Sections 7.9.3 14.2.11 and 14.2.1.4 of the SAR. hnprmement> in analytical capability have allowed a more renced anab sit of the control rod drop scisdent.

These techniques are described in a topical report (Reference 2) and two supplements (References 3 and 4). In additinn, a banked position withdrawal sequence descrj bed in Reference 5 has bocn developed to further reduce incrementnl rod wortha. Method and huuJ u for the rod drop accident at alysca are documented in Reference 3.

By using the analytical modds dewribed in those reports toupled with conservative or worst. case input parameters, it has been deterrmned that for power levels less thanM of rated pomer, the l

speci6cd krnit on in.equence control rod or control rod segment worths wdl limit the peak fuel enthalpy to less than 280 cal /t. Above20% power even single operator errors cannot resuh in out of sequence control rod worths which are sumeient to reach a peak fuel enthalpy of 280 cal E/

]

should a postulated control rod drop accident occur.

The following paramotors and worst-case assumptions have been utilir.ed peak fuel in the analysis enthalpy.

to determine compliance with the 280 cal /cm conformance to the limitin6 parameters,Each core reload will be analyzed to show

a. an interassembly local peaking factor (Referenco 6).

. Amendment No. 51 3 3/h.3-8 1

t

• _ _. . . . __ _ .~

QUAD CITIES DPR-30

b. the delayed neutron fraction chosen for the bounding reactivity curve l

c. a beginning-of-life Doppler reac tivity feedback d, scram times slower than the Technical Specification rod scram insertion ra te (Section 3 3.c.1)

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

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

g. the. moderator temperature at which criticality occurs In must cases the worth of insequence rods or rod segments in con, junction l with the actual values of the other important accident analyais parameters described above, would most likely result in a peak fuel enthalpy aub-stantially less than 280 cal /g design limit. I i Should a control 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 offsite 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.

The rod worth minimiter provides automatic supervision to assure that out of sequence control rods will not be withdrawn or inserted; ie., it limits operator deviations from planned withdrawal sequences (reference SAR Section 7.9). It serves as a backup to procedural control of control rod j 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 l 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 scram function, it does provide the operator with a visual indication of neutron level. This is needed for knowledgeable and efficient reactor startup at low neutron levels. The consequences of reactivity accidents are functions of the initial 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 monitor the 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 channels are provided, and one of these may be bypassed from the console for maintenance and/or testing. Tripping of one of the channels will block erroneous rod withdrawal soon enough to prevent fuel damage. This system backs up the operator, who withdraws control rods according to a written sequence. The specified restrictions with one channel out of service conservatively assure tnat fuel damage will not occur due to rod withdrawal 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 or more fuel rnds wi th MCPh's less than the MCPR fuel cladding integrity safety limit.During use orsuch patterns.

it is judged that testir.g of the RBM system to assure its operability prior to withdrawal of such rods will assure that improper withdrawal does not occur. !:is the responsibility of the Nuclear Engineer to identify these limiting patterns and the designated rods either when the patterns are initially established or as they develop due to the occurrence ofinoperable control rods in other than limiting patterns.

51 3 3A 3-9 Amendment No.

l

QUAD CITIES DPR-30 C. Scram Insertion Timea The control rod system is analy::ed to bring One reacter suberitical at a rate fast enougn to prevent fuel damage, i.e., to Orevent tne X"?E from hecoming less than the fuel cladding integrity safety limit.

' Analysis of the limiting power :ransient snews that :ne negative reactivity rates resulting frem the scram with the average resp 0:.3e f all the drives as civen in the above specification, provide tne require; protection, and MCfR remains greater than the fuel cladding intepity ,

safety limit.

The minimum amount of reactivity to be inserted during a scram is controlled by permitting r.o more than 10% of the operable rods to have long scram times. In the analytical treatment of the transients,390 milliseconds are allowed between a neutron sensor 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 mithseconds. Approu imately 70 milliseconds after neutron flux reaches the trip point, the pilot scram valve solenoid deenergires. Approumately 200 millisecondslater, control rod motion begins The time to deenerf ize the pilot valve scram >olenoids is measured during the cabbratmn tests required by Speci6 cation 41.The 200 millisceands are included in the allowable scram insertion times speci6cd in Speci6 cation 3.3.C.

The scram times for all control rods will be determmed at the time of each refuchng outage. A representative sample of control rods will be scram tested following a shutdown.

Scram times of new drives are approximately 2.5 to 3 seconds; lower rates of change in scram times following initial plant operation at power are expected.

provides reasonable assurance of detection of slow drives before system deterioration beyond the limits of Speci6 cation 3.3.C. The program was developed on the basis of the statistical approach outlined below and judgment.

The history of drive perforrcance accumulated to date indicates tL the 90% insertion times of new and overhauled drives approximate a normal distribution about the mean which tends to become skewed toward longer scram times as operating time is accurnulated 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 scram performance of the drives surrounding a drive exceeding the expected range of scram performance will detect local variations 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 surveillance of possibic anomalcus performance.

The numerical values assigned to the predicted scram performance are based on the analysis of the Dresden 2 startup data and of data from other IMit's 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 indicatmn of a systematic problem with control roJ drives, especially if the number of drives exhibiting such scram times exceeds eight, the allowahle number ofinoperable rods.

3.3 / 4.3 -10 Amendment No. 51

\_-__-_____-_-____________

QllAD-CITIES DPR.10 D. Control Rod Accumlators The basis for this specification was not described in the SAR and is therefore presented in its entirety.

Requiring no more than one inoperable accumulator in any nine. rod square array is based on a series of XY PDQ 4 quarter cote calculations of a cold clean core. The worst case in a nine rod withdrawal sequence resuhed in a k,, < l.0. Other repeating rod sequences with more rods withdrawn resulted in k, > l.0. As reactor pressures in excess of 800 psig, even those control rods with inoperable accumulators will be able to meet requ red scram insertion times due to the action of reactor pressure.

In addition, they may be normally in etted using the control rod drive hydraulic system. Procedural control will assure that control rods with inoperable accumulators will be spaced in a one in.nine array rather than grouped together.

E. Reactivity Anomalica.

During each fuel cycle. excess operating reactivity varies as fuel depletes and as any burnable poison in supplementary controlis burned. The rnagnitude of this excess reactivity may be inferred from the critical rod configuration. As fuel burnup progresses, anomalous behavior in the excess reactivity may be detected by comparison of the critical rod pattern selected base states to the predicted rod inventory at that state.

Power operating base conditions provide the most sensitive and directly interpretable data relative to core reactivity Furthermore, using power operating base conditions permits frequent reactivity comparisons.

Requiring a reactivity comparison at the specified frequency assures that a comparison will be made before the core reactivity change exceeds 1% ak. Deviations in core reactivity greater than 1% ok are not expected and require thorough evaluation. A 1% reactivity limit is considered safe.since an insertion of the reactivity into the core would not lead to transients exceeding design conditions of the reactor system.

F. Econoinie Generation Control Sptem Operation of the facility with the economic generation control system (EGC)(automatic flow control) is limited to the range of 65% to 100% of rated core flow. in this flow range and with reactor power above 20%, the reactor could safely tolerate a rate of change ofload of 8 MWe/sec (reference SAR Section 7.3.5 ).

Limits within the EGC and the flow control system prevent rates of change greater than approximately 4 MWe/sec. When EGC is in operation, this fact will be indicated on the main control room console. The results ofinitial testing will be provided to the NRC before the onset of routine operation with EGC.

References

1. " Generic Relmd c'ue l A pplica t,lon " , NEDE-24011-P-A" l

2. C. J. Paone, R. C.Stirn, and J. A. Wooley,* Rod Drop Accident Analysis for Large BWR Y GE Topical Report i NEDO 10527, Manh 1972.

3. C. J. Panne. R. C.Stirn and R. M. Young. ' Rod Drop Accident Analysis for Large BWR's'. Supplement 1. l GE Topical Report NEDO 1052 7. July 1972.

4. J. M. Haun, C. J. Paone, and R. C. Stirn. ' Rod Drop Accident Analysis for Large BWR's, Addendum 2. I Exposed Cores,' Supplement 2. GE Topical Report NEDO 10527, January 1973.

5. C . J. Paone, "%nked position withdrawal sequence," Licensing topical. Report NEDO-21231, January, 1977

6. ' To include the ' power spike effect caused by gaps between fuel pelleta. l

• Approved revision rmmber a t t.imo reload fuel anatyaua are po r N etnul . l Arnendment flo. 51 3. V4. 3-11

QUAD CITIES D PR-30 3.3 LD1ITI:iG 00:iDITION3 FOR OPERATIOM BASES 1

A. The decign ob.jective of toe atandby liquid control yster IJ to l ccvile the capaoility ur or!.nging the reactor fr/n full mwer to a cold, xenon-free shutdown assuming that none of the withdrawn control rods can ce inserted. To meet thia ab.iee tive, the liquid control syste:n is designed to in, ject a quantity of Doron which i produce; a concentration of no less than 600 ppm of boron in the j reactor core in approximately 90 to 120 minutes with imperfect mixinr. A boran concentration of 000 ppm in %e reactor care is required to oring the reactor from full pcwer to 3% Ak or mare }

aucaritical condition considering the hot to cold reactivity swit. ;,

xenon poisoning and an additional margin in the reactor core for imperfect miune of the i chemical so!ution in the reactor water. A normal quantity of 3470 gallons of solution having a 13M sodium pentaborate concentration is required to meet this shutdown requirement The time requirement (90 to 120 minutes) for insertion of the boron solution was selected to override the rate of teactivity insertion due to cooldown of the reacter following the xenon penon peak. For a required pumping rate of 39 gpm, the maximum storage volume of the beton solution is established as 4875 gallons (195 gallons are contained below the pump suction and, therefore cannot be inserted).

Boron concentration, solution temperature. and volume are checked on a frequency to assure a high reliability of operation of the system should it ever be required. Experience with pump operabiltty indicates that monthly testing is adequate to detect if failures have occurred.

The only practical time to test the standby liquid control system is during a refueling outage and by initiation from local stations. Components of the sy stem are checked periodically as described above and make a functional test of the entire system on a frequency of less than once each refueling outage unnecessary. A test of explosive charges from one manufacturing batch is made to assure that the charges are satisfactory. A continual check of the firing circuit continuity is provided by pilot lights in the control room.

B. Only one of the two standby liquid control pumping circuits is needed for proper operation of the system.

If one pumping circuit is found to be inoperable, there is no immediate ireat to shutdown capability, and reactor operation may continue while repairs are being made. Assurance that the remaining 9 stem will perform its intended function and that the reliability of the system is good n obt.nned by demonstrating operation of the pump in the operable circuit at least once daily. A reliability analysis indicates that the plant can be operated safely in this manner for 7 days.

C. The solution saturation temperature of 13% sodium pentaborate, by weight,is 59' F. The solution shall be kept at least 10' F above the saturation temperature to guard against boron precipitation 't he 10' F margin is included in Figure 3.31. Temperature and liquid level alarms for the system are annunciated in the control room.

Pump operability is checked on a frequency to assure a high reliability of operation of the system should it ever be required.

Once the solution has been made up, boron concentration will not vary unless more boron or rnore water is added. Level indication and alarm indicate whether the solution volume has changed, which micht indicate a possible solution concentration change. Considering these factors, the test interval has been established.

l Amendment No. 51 3.4/4.4-3

QU A D-CITIF.S Di'R-30 cycle by assuring that water can be run thiough the dr.un lines and actuating the air operated vahes by operation of the following sensors:

1) fou or air

2) etjuipment drain sump hi,oh les el

3) v.mit hi;'h Ics el

d. The rondenser pn 5.f oot trip to -

tuns lor e.wh channel shall be checked once a month A logic system functional test shall be per-formcd during each refueling ou tage.

I. Awrage Planar LilGR I. Average Planar LHGR During steady. state power operation, the averrge linear heat generation rate (APLliGR) of all the Caily during steady 5 ta te opera *' ~

rods in any fuel assembly, as a function of avetage a,bCve 2 % rated Oner nal pc.[fo r planar exposure, at any axial location, shall not the a'/erage planar LHOR sha"'l '

exceed the maximum average planar LIIGR be determined.

shown in Figure 3.51, if at any time j during operation it is determined by norinal sur-veillance that the limiting value for APLilGR is j, toegg tygg being exceeded. action shall be miriated within 15 minutes to restore operation to within the pre. Daily during steady state power operation scribed limits. If the APLilGR is not returned in above 25e6 of rated thermal power, the local within the prescribed hmits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, the LHGR shall be determined.

reactor shall be brought to the cold shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Surveillance and corresponding action shall continue until reactor operation is within the presenhed limits.

J. Local LilGR During steady state power operation, the linear heat generation rate (L!lGR) of any rod in any fuel essembly at any axial location shall not exceed the maximum allowable L}lGR If at any time during operation it is determined by normal surveillance that the h'mting value for LilGR is being exceeded, action shall be initiated withm 15 minutes to restore operation to withm the pre-scribed limits. If the LilGR is not returned to Amendment No. 51 3* "

i

l QUAD CITIES DPR-30 within the prescribed limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, the reactor shall be bruught to the cold shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Surveillance and cor.

responding action shall cont:nue until reactor operation is withm the prescribed hmits.

Maximum allowable LHGR for all 8XC fuel tl/ pes is 13 A KWf t.

For 7X7 and mixed oxide fuel, the riaximum allowable L!!GR 1a as follows:

LHG R,,,,, <l.!!G R,, 1 -t 3 P/ P ),,, ,,( L / L , )

w here-LilG R, = , design LilG R

= 17. 5 k W/f t.

( A P/ P ),,,,,, = masimum power spiking penalty

= .035 initial core fuel

= .029 reload 1. 7 x 7 fuel I

= .028 reload 1,7 x 7 mixed oxide fuel L, = total core length

= 12 feet L = Axial distance from bottom of enre K. Minimum Critical Power Ratio (MCPR) K. Minirnum Critical Pimer Ratio (N!CPR)

During steady 4 tate operation MCPR shall be Th- MCPR shall be determined d.nly during greater than or equal to steady-state power operation .ibos e 2.; 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 surveillance that the limiting value for MCPR is being exceeded, action shall be imtiated within 15 minutes to restore operation to within the prescribed Irmits.

If the steady. state MCPR is not returned to within the prescribed Itmits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, the reactor ,

shall be brought to the cold shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Surveillance and corresponding action shall continue until reactor operation is within the prescribed limits. For core flows other than rated, these norninal values of MCPR shall be increased by a factor of kg where kg is as shown in Figure 3.5.2.

l

. l Amendment No. 51 3.5/4.5-10 l

QUAD-CITTS DPR-30 diesel generaters. All of these systems have bee". sized to perform their intended function considering the simultaneous operation of both units.

These technical specinutions contain only a single reference to the operability and surveillance requirements for the shared safety related TCJtures of each plant. The level of operahility for one unit must be maintained independently of the status of the other. For example, a diesel (1/2 diesel) which is shared between Units I and 2 would hase to be operable for continuing Unit 1 operation even if Unit 2 were in a cold shutdo.vn condition and needed no diesel power.

Speci6 canon 3 5 F.3 provides that should this occur, no work will be performed which could preclude adequate emergency cooling capability being avadab:e. Work is prohibited unless it is in accordance with specified procedures which limit the period that the control rod drive housing is open and assures that the worst posab!e Inss of coolant resulting from the work will not result in uncoverity the reactor core Thus, this specification assures adequate core cochng. Specification 3 9 must be consulted to determine other requirements for the diesel generator.

G. Maintenance of Filled Discharge Pipe If the discharge piping of the core spray, LPCI mode of the RHR, HPC, and RCIC are not 611ed, a water hammer can develop in this piping, threatening system damage and thus the availability of emergency cooling systems when the pump and/or pumps are started. An analysis has been done which shows that if a water hammer were to occur at the time emergency cooling was required, the systems would soll perform their design function. However, to minimite damage to the discharge systems and to ensure added margin in the operation of these systems, this technical specification requires the discharge lines to be filled whenever the system is in an operable condition.

Specincation 3.5.F 4 provides assurance that an adequate supply of coolant water is immediately available to the low-pressure core cooling systems and that the core will remain covered in the event of a loss of coolant accident while the reactor is depressurized with the head removed.

H. Condensate Pump Room Flood Protection See Speci6 cation 3.5.H.

1. Average Planar LHGR This specification assures that the peak cladding temperature following a postulated design basis loss of coolant accident will not exceed the 2200' F limit speci6ed in 10 CFR 50 Appendix K considering the postulated efTects of fuel pellet densification.

The peak claddinc temperature following a postulated loss-of coolant accident is primarily a function of the average LHGR of all the rods in a fuel assembly at any axial location and is only secondarily dependent on the rod to rod power distribution within a fuel assembly. Since expected local variations in power distribution within a fuel assembly aflect the calculated peak cladding temperature by less than 120' F relative to the peak temperature for a typical fuel design, the limit on the average planar LHGR is sufficient to assure that calculated temperatures are below the 10 CFR 50 Appendix K limit.

The maximum average planar 1. HOR's shown in Figure 3.5-1 are based on calculations employing the medcls described in Reference 2. Power operation with LHOR's at or below those shown in Figure 3 51 l assures that the peak dadding temperature following a postulated loss-of. coolant accident will not cuecd the 2200' F hrnit lhese values represent linuts t'or operation to ensure conformance with 10 CFR 50 and Appendix K only if they are more hmiting than other design parameters.

The rnaximum average planar LHOR's plotted in Figure 3.51 at higher exposures result in a calculated peak cladding temperature er less than 2200

• F. However, the maximum average planar LHGR's are 3M 543 Arnendment flo. 51

a g g QUAD-CITIES DPR-30 shown on Figure 15-1. as limits because conformance calculations have not been performed to justify operation at 1.HGR's in excess of those shown.

J. Local LHGR This specification assures that the maximt.ru linear heat generation rate in any rod is less than the design linear heat. generation rate even if fuel pellet densification is postulated. The power spike penalty

,13 discussed in Peference 2 and assumes a linearly increasing variation in aval l gaps between core bottom and top and assure with 95% confidence that no more than one fuci rod exceeds the design LHGR due to power spiking.

K. Minimum Critical Power Ratio (MCPRI ne steady state values for MCPR specified in this specification were selected to provide margin to acconuno-date transients and uncertainties in monitoring the core operating state as well as uncertaintres in the critical power correlation itself. These values also assure that operation wi!! be such that the initial condition assumed for the LOCA analysis plus two percent for uncertainty is sa tisfied . For l any of the special set of transients or disturbances caused by sing!e operator error or single equipment malfunction,it is required that des:gn analyses initialized at this steady state operating hmit yield a MCPR of not less than that specified in Speci0 cation 1.1.A at any time during the transient, assuming instrument trip settings given in Spec 10 cation 2.1. For analysis of the thermal consequences of these transients, the value of MCPR s tated in this specification for the limiting condition of operation bounds the initial value of MCPR assumed to exist prior to the initiation of the transjents.

This initial condition, which is used in the transient analyses, will pre-clude' violation of the fuel cladding integrity safety limit. Adsumptions and methods used in calculating the required steady state MCPR limit for each reload cycle are documented in Reference 2. The results apply with increased conserva tism while operating with MCPR's greater than specified.

The most limiting transients with respect to MCPR are generally:

a) Rod withdrawai rrror b) Imad rejection or Turbine Trip without bypass c) Loss of feedwater heater deve ra factors influence which of these transients results in the l a rges t reduction in critical power ratio such as the specific fuel loading, ex-posure, and fuel type. The current cycles reload licensing analyses spec- ]

ifies the limiting transients for-a given exposure increment for each fuel type. . The values specified as the Limiting Condition of Operation are con-servatively chosen to bound the most restrictive over the entire cycle for l ewh ruoi tyne.

l.

Amendment No. 51 3, w ,5_14

QUAD CITIES DPR 30 For core Dow rates less than rated, the s:cady state MCPR is increased by the fermeia given in the specifi-cation. 3:s anures that the MCPR will be maintained greater than that specified in Specincatmn 1.1.A esen in the event that the motor generator sct speed controller causes the scoop tube positioner for the Gud cuap to move to the rnnimum speed position.

References

" Lass-of-0colan* Analysis Report for Dresden tJnits P. 2 ard auad Cittec

, , r -t. m,

'J n t > 1, 2 'luclear Pc.rer Stations," ..w- .n'.

.y .- , .

'. "Jeneric Relo r! Fuel Applica tien, " IIEDE-24011-P-A" 3 "

. . Jacobs and F. W. Ma rrio t t , GE Topical Report APED C L, J" Guidelines f ar Letermininc Jafe Tes t Intervals 1;r Er.gi .ee r ?1 Jafer,uarda,' Acril, 1969 ena 1". mar.d

. ec t

+

Approved revision at time of plant operation.

Approved performed.

revision number at time reload fuel analyses are

)

1 3 5/4.5-14a Amendment No. 51 l

l

• DPR-30

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Amendment No. 51

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MAX 1uuM AVE RAOC TL *J4 AR LIM AR HE A T GE M 4 A TION H ATE Waft.HCR) .

Amendment No. 51

Quad Cities LPB_30

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. lfA#.LEJvi AVE ftAG E FL." *. A R LIN E AR H E AT GE N E 9 A WW R ATE (M.APLH GN Vf P.5US PLAN A R AVE R.A GE E.XPCSURE l

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Amendment No. 51 D 3'M l (Cheet b of h)

I i

--. - y w -ia.-- .,.a-- +w .+.-.

4 9 Qt! A D-CITIES DPR-30 Shou!J the switthes at feveh (a) and Ih) fail or the operator fail to trip the circulating water pumps on alarm at level (h), the actuanon of either level switch pair at level (c) shall trip the citculatine water pumps attomatically and alarm m the control room These redundant level switch pairs at level (c) are designed and trataUed to !HT 279, 'Critetia for Nuclear Power Mant Protection Sptems' As the circulating water pumps are tripped, either manually or automatically, at level (c) of 5 feet, the maximum water level reached in the condenser pit due to pumping will he at elevauon $68 feet 6 inches elevation (10 feet above condenser pit Coor elevation 558 feet 6 inches: 5 feet plus an addnional 5 feet attributed to pump coastdown).

In order to prevent the RilR seruce water pump motors and dicsci generator coo!ing water pump motors from oserheating, a sault cooler it supplied for each pump 1 ach vault coaler is designed to maintain the vault at a maximum 105' F temperature durmg operanon of its respective pump For examp!c. if diesel genetator coohnr water pump 1/2 3903 starts,its cooler also starts and m.untains the sault at 105" F by removing heat sapphed to the vault by the motor of pump 1/2 A903. If. at the sarne time that pump 1/2 3903 is in operanon. RHR sen ne water pump IC starts,its cooler will also start and compensate for the added heat supplied to the sault by the IC pump motor keeping the vault at 105' F.

Each of the coolers is supplied with cooling water from its respective pump's discharge line. After the water has been passed through the cooler it returns to its respective pump's suction line. In this way the vault coolers are supplied with cooling water totally inside the vault. The cooling water quantity needed for each cooler is approximately 19 to 59 of the design now of the pumps so that the recirculation of this small amount of heated water will not affect pump or cooler operation.

Operation of the fans and coolers is required during shutdown and thus additional surveillance is not required Watertight vaults for the ECCS pumps in the reactor building are tested in essentially the same manner and frequency as described for the condenser pump room vaultt Veri 6 cation that access doors to each vault are closed following entrance by personnel is covered by station operating procedures.

The LHGR shall be checked daily to determine if fuel burnup or control rod movement has caused changes in power distribution. Since changes due to burnup are slow and only a few control rods are moved daily, a daily check of power distribanon is adequate Ascrage Planar LilGR At core thermal pow er levels less than or euual to 25% operating plant experience and thermal hydraulic analpes indica:e that the resulung average planar LilGR is below the maximum average planar LilGR by a consideraNe margin: therefore, evaulation of the average planar LilGR helow this power lesel is not necessary. The daily requirement for calculating average planar LilGR .aove 25% rated thermal power is su0icient, since power distribunon shifts are slow when there have not been significant power or control rod changes.

l.ocal LilGR The LilGR as a function of core height shall be checked daily during reactor operation at greater than or equal to 25% power to determine if fuel burnup or control rod movement has caused changes in power distribution. A limiting IBGR value is precluded by a considerable margin when employing any permiscible control rod pattern below 25% rated thermal power.

Minimum Critical Power Ratio (MCPR)

At core thermal power levek less than or equal to 25% the reactor will be operating at minimum recirculation l pump speed and the moderator void content will he very small. For all designated control rod patterns which may be employed at this point. operating plant esperience and thermal hydraulic analysis indicate that the resulting MCPR value is in excess of requirements by a considerable margin. With this low void content, any inadvertent core now increase would only place operation in a more conservative mode relative to MCPR.

Amendment No. 51 3.s/.: s.17 i

l

, ()ll Al) ( l l'IEN I)PI(- 10

2. lioth the sump and an urnpimpi y s-tems .sha!! be operah:r Juring rea ter powcr operainen l'iorn .nid ane: tl e date that one of these sy sienn e made or four;J to be inoperab!c for any re.o ton, reactor power opcf ilion is pel-minible only Jurine the sucu cJmy 7 days.

3. If the conditions m I os 2 abme va.

not be met. an orderls sinnJou i shall be ininated .md the re.n tot shall he m a cold shutdow n condirnm u nion 21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br /> I .Naftt) arid llfliCI Y.th C% l. ,N.ifel) 3 HI l{Cliol Y.shrs

1. Prior to reacter st iriup for powet op A nununum of I // of all ulety uh es sh.ill be eration, durine reador power operat. bench chesLed or replat ed uuh a hent h I ing conditions.and w benever the reau checLed sah e cat h reluchny oulare ibe pop  !

tor coolant prenure n picater than 90 ping potut of the urety sahes sh.dl be sci .n j psig and temperature greater than follow s-j 320 ' f . all nine of the vfet,v sahes j shall be operahle. T he solenoirl- Number of Valves Serpomt (pug) activated pressure s ak es shall be oper-g able as required by Specitica tion .,

3 5J) ~

2 1250

2. If Specification 3 f,T.. I is not met, the 4 12 <,0 reactor shall remain shut dow n unnl the condition n i or rected oi. if in The allowable scipomt citor for each uke is operauon. an orderly shuidow n shall i1%

be initiated and the scacior coolant prenure and teinperature shall be All rebel. valces shall be checked toi set Ises below 90 psip and 320 I willon 24 sum each aluchny outare Ihe set piewums hours shall be.

Number of l'alves Serpomt (psig)

I 1 lise" {

2 s 11.10 2 < l 13 A "Ta rget Rock tombniation salety / rchef vahr F. Structural Integrity F. Structural Intep!!y

'lhe structural intepity of the prinury system The nonder.tinctive impections Inted m Table boundary shall be maintained at the lesel required 4.61 shall he pertooned as specihed m accor-by the ASMl: lloder and Pressuie Ves.el Code, dance with Sntion XI of the ASML lioiler and Section XI. " Rules for Inwsuce Inspection ef Pressure Vessel Code, 1971 I dition, Summer Nuclear Power ' Plant Coinponent s", 1974 1971 Addenda, the results obtauied from com-

, I dition. Sainmer P)75 AJJenda ( ASM.: Code phar,ee with this speathation will he culu.ited i Sedion XI). after 5 years an.1 the conchisions wdl be re-i t

viewed with the NitC.

l Amendment No. 51 3.r.f a <, . .I

-