ML20062H831
| ML20062H831 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 09/02/1980 |
| From: | COMMONWEALTH EDISON CO. |
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| Shared Package | |
| ML20062H829 | List: |
| References | |
| NUDOCS 8009080341 | |
| Download: ML20062H831 (47) | |
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. ENCLOSURE I Quad Cities Station Unit Proposed Changes to Appendix A, Technical Spec 1fications To Facility Operating License DPR-29 Revised 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 1.1/2.1-4 3.3/4.3-10 1.1/2.1-5 3.3/4.3-11 1.1/2.1-6 3.4/4.4-3 1.1/2.1-7 3.5/4.5-9 1.1/2.1-8 3.5/4.5-10 1.1/2.1-9 3.5/4.5-11 1.1/2.1-10 3.5/4.5-14 1.1/2.1-11 3.5/4.5-15 1.2/2.2-2 3.5/4.5-18 1.2/2.2-3 Figure 3.5-1, Sheets 1 3.1/4.1-1 through 6 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 3.7/4.2-8 3.2/4.2-14 3.2/4.2-15 New pages: 1.1/2.1-2a Deleted pages: Figure 2.1-2 8 0 09080 '3 L{ {
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, QUAD.CITITS DPR-29 1
1.1/2.1 FUEL CLADDING INTi{GitITY
. SAIITY IJMIT I.lMITING SArl'.TY SWlf.M Sr.TilNG Applicability: Appilcability:
The sareiy limits establishtti to preserve the fuel The limiting safety sysicm settings appl 3 to trip cladding integrity apply to ihme vJri.slaIes whtCh settings of the instruments and devices whith are snonitor tae fuct thermal behmor provided to paesent the fuct cladding integrity eefety hmits from being caceccled.
Objectiee: Objectiie:
The objective of the safety limits is to estehtish The objective of t!ic limiting safety systern settines limits b: low which the integrity of the fuct cladding is to denne the level of the procos variabtes at v.hich is p eserved.
automatic proicctive action n initiated to prevent
( the fuel eladdmg integrity safety limits frorn being szeceded.
SPECIFICATIONS O
A. Reerior rnssuse > S00 rig and Core llow A. Neutron I' lum Trip Settint.s l > 10% of Rsted ne limitinc ufety systern trip settings shall be 9he eXISteDCe of a mim1Pml as speci$ed be?ow:
critical power ratio (fCPR) 1. APRM rlum scram Trip Settin; tnun less than 1.07 shall constitute gode) violation of the fuel clad- when the reactnr mode switch is in it:e ding integrity safety limit, g,o pos;iion, ii e Ai.nu tlus sciam setting, shall I.e as shewn in I-agune 2.l.1 and shall lic:
- 3. Core Thermal Power Li nit (Resetor Pre %ure s 800 psi:) S$(.65WD + $5) l with a nominwm setpoint of 120% for When the reactor pressure is s 500 psig or " '" * #
core Sw h less than 10% of rated, the core l" I''"'
therrnal pov er shall not exceed 25G cf rated
. thermal power. where:
- S "
Seiiint ni Percent of rated C. romer Transient power I. The neutron out shall not esned the percent of drive now te- \
scram settinr est.:blished in Spectrica-WD psra w proauce a rated cos e tson 2.l A for longer th n 1.5 seconds tsow or 9tsintition ab/hr. In as indicaico by the process cornputer. the event et c pen et ton with a maximam fract ion of limit arr)
. 2. When the process computer is out of pp.or donetty twt.rri) greatee service. Ilu,s sikty hmit th.ill ta as- than the treetsos or cated sumed to be escudeel if the neution power (fyl-), the f.ctting shall Out eacceds Ilic strann wtiiry c.I.il,, M modified at f ol lW.
d fished by Sg. citie iion 2.1.A and a control ruJ sram does r.ot etcur. a 6 (.61st, + 551 [1,,, prt.m.1,7f P]
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QUAD-CITIF.S DIH-29 ,
O H. Umiting Conditions for Operation (LCO) The limiting conditions for operation specify the minimum acceptable levels of system performance nesxssary to assure safe startup and operation of the When these conditions are met, the plant can be operated safely and abnormal situations can be safely controlled.
L Limiting Safety System Setsing(LSSS) -The limiting safety system settings are settings ofs ins tion which initiate the automatic protective action at a level such that the safety limits will not be exceeded. The region between the safety limit and these settings represents margin, with normal operation lying below these settings. The margin has been established sd that with pr the instrumentation, the safety limits will never be exceeded. ,
K. 14:lc System Funerlonal Test - A logic system functional test means a test of all relays a logic circuit from sensor to activated device to ensure all components are operable pe Where possihte. action will go to completion; i.e.. pumps will be started and valves opened 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: *
- l. Shutdown -In this position, a reactor scram is initiated. power to the control rod drive > is rem and the reactor protection trip systems have been deenergized for 10 seconds prior to pes g manual reset.
- 2. Refuel - In this position, inter!ocks are established so that one control rod only may be withd when flux amplifiers are set at the proper sensitivity level and the refueling crane is not ove reactor. Also, the trips from the turbine control valves. turbine stop valves main steam isolatio valves, and condenser vacuum are bypassed. If the refueling crane is over the reactor, all rods m be fully inserted and none can be withdrawn.
- 3. Startup/ Hot Standby - In this position,the reactor protection scram irip . initiated by co vacuum and main steamline isolation valve closure. are bypassed, the low pressuie main steatn isolation valve closure trip is bypassed.and the reactor protection system is energi7ed. with APRM neutron monitoring system trips and control rod withdrawal interlocks in service.
- 4. Run - In this position the reactor system preaure is at or above 850 psig.and the reactor system is energized.with APRM protection and RMB interlocks in service (exclud flux scram).
M. Operable A system or component shall be considered operable when it is capable intended function in its required manner. .
l N. Operating Operating means that a system or component is performing its intended
! required manner.
O. Operating Cycle Interval between the end of one refueling outage for a particular the next subsequent refueling outage for the same unit.
P. Primary Containment Integrity - Primary containment integrity means that the drywell and pressure suppression (hamber are intact and all of the following conditions are sainfied:
g 1. All manual containment iso!.stion valves on lines connecting to the reactor coolant system or containmen'. which are not required to be open during accident conditions are closed.
1.0-2
QUAI)-CITIF.S DPR-29 v
Y. Sheidnwn The reactor is in a shutdown condition when the reactor mode switch is in the Shutdow n posmon and no core alterations are beint performed. .
I. Hot Shutdown means conditions as above with reactor coolant temperature greater than 212' F.
- 2. Cold Shutdown means conditions as aime. with reactor coolant temperature equal to or leu than 212 F.
- 2. Simulated Automatic Aesuasion - Simulated automatic actuation means applying a simulated ugnal to the sensor to actuate the circuit in question.
- 88. Transition Be[iling Transition boiling rneans the boiling regime between nucleate and film boiling.
Transition boiling 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 assembly power which causes some point in the auembly to experience transition boiling to the auembly power at the reactor condmon ofinterest as calculated by apphcation of the GEXL correlation (reference NEDO.109510 DD. Minirnum Critical Pon er Ratio (MCPR) - The minimum incore critical pcwcr ratio corresponding to the most limiting fuel assembly in the core.
EE. Suncillance Intenal - Each surveillance requirement shall be performed within the specified surveil.
Iance interval with:
- a. A maximum a!!cwahle 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.
FF. Fraction of Limiting Power Density (FLPD) - he fraction of limiting power density is the ratio of the lin.ar heat generation rate (LEGR) existing at a given location to the design LHGR for that bundle type.
GG .. Maximun 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 (FRP) - he fraction of rated power is the ratio of core thermal power to rated thertal power of 2511 MWth.
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QUAD-CITIES '
DPR-29 O '
where:
D. Reactor Water leel (Shutdown Condition)
FRP = "
Whenever the reactor is in the shut- te p down condition with irradiated fuci (2511 MWt) in the reactor vessel, the water level shall not be less than that MFLPD = maximum fraction of corresponding to 12 inches above the limiting power dens-l top of the active fuel
- when it is ity where the lirait-seated in the core. ing power density for each bundle is
- Top of active fuel is defined to be the design linear -
360 inches abovo vessel zero (See heat generation rate Bases 3.2), for that bundle.
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The ratio of FRP/MFLPD shall be set equal to 1.0 unless the actu-al operating value is less than 1.0 in which case the actual operating value will be used.
This adjustment may also be performed l by increasing the APRM gain by the l inverse ratio, MFLPD/FRP, which accomplishes the same degree of pro-tection as reducing the trip setting by FRP/MFLPD.
- 2. APRM Flux Scran Trip Setting (Re.
fueling or Startup and fict Standby Mode)
When the reactor mode switch is in the Refuel or Startup flot Standby posi- !
tion, the APR.\1 scram shall be set at
' less than or equal to 15% of rated ,
neutron flux.
' 'Ihe IRM flux scram wtting shall be set at less than or equal to 120/125 of full scale.
- 4. When the reactor mode switch is in the j startup or run position, the reactor shall
- not be operated in the natural circula.
tion flow mode.
B. APRM Rod Block Setting The APRM rod block setting shall be as shown
' in Figure 2.1 1 and shall be:
S s (.65V(p+ 43)
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QUAD-CITIES nPR-29 The definitions used above for the APRM scram trip apply. In the event of oper-ation with a maximum fraction limiting powcr density (IELPD) greater than the '
fraction of rated power (FRP), the se.tting shall be modified as follows:
FRP S 6 (.65tig + 43) FJLPD The definitions used above for the APRM ,
scram trip apply.
The ratio of FRP to PJLPD 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.
This may also be performed by increasing the APRM gain by the inverse ratio, MFLPD/FRP, ,
which accomplishes the same degree of pro- ,
tection as reducing the trip setting by FRP/MFLPD.
C. Reactor low water level scram setting
! shall be 144 inches above the top of the
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" active fuel
- at normal operating condi-tions.
D. Reactor low water level Eces initiation shall bo 84 inches (+4 inches /-0 inch) above the top of the active fuel
- at .
2 normal operating conditions.
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E. Turbine stop valve scram shall be s 10% valve 4
closure from full open.
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I F. Turbine control valve fast closure scram shall initiate upon actuation of the fast closure sole.
noid valves which trip the turbine contro!
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valves.
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I G. Main steamline isolation valve closure scram shall be s 10% valve closure from full open.
H. Main steamline low-pressure initiation of main l steamline isolation valve closure shall be 2: 850 psig.
- beTop of ac-tive fuel is defined to 3 60 inches above vessel zero (See Bases 3 2)
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QUAD-CITIES OPR-29 1.1 SA WTY LIMIT BASIS The fuel cladding integrity limit is set such that no calculated fuel de.ioage would necur as a recu1* of an abnormal operational trsnstent. Decause fuel damage is not directly cbacavable, a stup-back arnocch is used to establish a safety Itnit such that the manicum critical power r.itio (PCrn) cladding integrity saf ety Itnat.rcht > the fuel cledding integrity safety linit represents a conservativeas no less than3 the fut1l margin relative to the conditions required to maintalra fuel cladding integrity.
The fuel The cladding integrity is one of this of thebars cladding physical ier inbaralers relatsti which separate radioactive materials fros the cavirons.
to its relative f aredom from perfos etions or cracking.
Although sorne corrosion or use-related cracking nay occur during the life Cf the cladding, fassacn proJuct migration from this source is inerenentally currulative and continuou=1y escanurable. A uel claddang per-forations, however, can result frers thermal stsuces which occur fre;s a cactor operatsen signific ntly above design conditions and the protection systen cafsty settings. While fission pro;act Mgratien f ec16 clsodin; perforation is just cs measurable as that from vec.related cracking, the thermally c uted claddie.g perfos.
ations signal a threshold beyond wh'ich still greater thermal stresses swy cause gaout rather than nr.crt.vnt. ,
al cladding deterioration. Therefore, the fuel cladding safety limit is defined with raargin to the ccats- 1 tions which would produce onset of transation boiltng DtCPR o f 1.0) . Thcr.e conditionr. represent a regniti. ,
cant departure frun the condition intended by design for planned operation. Therefore, the fuel et ,eMing integrity safety limit is catablished such that no calculated fuel d mage sh;211 result f ror:1 ein
' abnornal operational transaent. Basic o' the values derived for this cafety limit f or each fuel type is documented in Reference 1.
A. Reactor Pressure > 800 peig and Core riow > lt% of mated Onset of transition boiling results in a decrease in host transfer frors the claddirig and thernfore
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elevated cladding temperrture and the possa' o ility of cladding f ailure. Ilowever, the entstence of critical p<.Ner, or boaling transition 16 not a directly observablu paren oter an an opsratiny reaet.
or, Therefore, the sargin to boilang transition is calculated frru plant opuratang parrmetras such
, as core power, core flov, feet; water tenperature, and core power dists abut sun. Tha margin for each (T -
fuel assembly is characterized by the critical power ratio (CrH), which as tha ratio of the 1.andle 1 -
power which would produce onset of transitaon boiling divided by the actual Lundle po.cr. The alnimura valuc of this ratio for any Insn21c in the core is thu minimum critical power ratio (ncrn).
It is assumed that the plant operation is controlled to the norsinal protectave netponta via tho
, instrumented variables (rigure 2.1-3).
The MCPR fuel cladding integrity safety limit has suf ficient con;ervatitas to assure that in the cver.tl of an abnormal operational transient initiated from the normal operating conh tion, more tnan 99.m of the fuel rods an the core are expected to avoid boiling transition. The marg s n beten !.CPit of 1.0 (oncet of transition boiling) and thc. safety limit, is derived from a detailed statistacat l l analysim considering all of the uncertainties in nionitoring the core operating state, sneluding uncertainty in the boiling transition correlation (see e.g., lec tere nc e 1). Because the boalang transition correlation is hated on a laree quantity of full-senlo data, there is a very high con.
fidence that operation of a fuel assembly at the condition of MCI'M - the fuct cladding integrity l safety limit would not produce boiling transation. a However, if boiling transition were to occur, cladding perforstion would not be expected. C1rdding temperatures would increase to appronanntely 1100*r, which is below the perforation temperature of the cladding raaterial. This has been verified try tests in the General r.lectric Test Reactor (Cf'ra) ,
where similar fuel operated above the critical heat flux for a significant pea-iod of tame (3C main-utes) without cladding perforation.
If reactor pressure should ever exceed 1400 psia during normal power operation (the limit of applicability of thu boiling transition correlation), it would be assumed that the fuel cladding integrity safety limit has been violated.
In addition to the boiling transition limit (MCPR) oparation is constrained to a maximum LH CRs17.5 kw/ft for 7 x 7 fuct and 13.4kw/f t for all 8x8 fuel types. This constraint is established by specification 3.s.3.
serain for abnormal to orovide operating adecuate safety init.ated
' transients margin to fro1 %m plastic high a power conditions. Specification 2.1.A.1 provides for equivalent safety margin for transients initiated from lower power con-ditions by adjusting the APRM flow-biased scram setting by the ratio of FRP/MFLPD. .
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Speci5 cation 3.5) established the LilGR maximum which cannot he exceeded 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,O flow)is Breater than 4.56 At low powers and flows thh pressure differential is maintained in the bypass region of the core. S the pressure drop in the bypass region is essentially all elevation head, the core pressure drop a powers and flows will always be greater than 4.56 psi. Analyses show that with a flow of 28 x bundle flow, bundle pressure drop is nearly independent of bundle power and has a value or3.5 psi. 'Ihu the bundle flow tvish a 4.56 psi driving head will be greater than 28 x 10' tb/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 btmdle would ha i
to be operating at 3.86 times the average powered bundle in order to achieve this bundle power. Th
' a core thermal power limit of 25% for reactor pressures below 800 psia is conservative.
> C. Power Tramient
'During transient operation the heat flux (thermal power to-water) would lag behind the neutron flux du to the inherent heat transler time comtaut of the fuel, which is 8 to 9 secomh. Abo. the limiting safety system scram settings are at values which wit! not allow the reactor to be oper.ited above the safety G during normal operanon or during other pla.it operating situations which have been analyzed in detail.
V In addition, control rod scrams are such 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 flux scram setting and a failure of the control rods to reduce flux 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 this scram setting is exc for longer than 1.5 seconds.
If the scram occurs suet that the neutron flux dwell time above the limiting safety system setting is less than 1.7 seconds, the safety limit will not be exceeded for normal turbine or generator trips, which are the most severe normal operating transients expected.These analyses show that even if the bypass system [
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 additional ritarain.
The computer provided'lias a sequence annunciation progrars which will indicate the sequence m wh scrams occur, such as neutron flux, pressure, etc. This program also indicates when 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 compnier information should not be available for any scram analysis, Specilication 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 c!Tect of decay heat. lf reactor water level should drop below the top of fuel during this time, the ability to cool the core is reduced. This reduction in core-cooling capability could lead to elevated cladding temperatures and cladding perforation.The core will he cooled suificiently to prevent dadding melting should the water level be reduce 4to two-thuds the coie height 1.si.thh ment of the ufety hinit at 12 inches abose the top of the fuel provides adesgo.pe marrin. 't his level u di 9 be contiuunusly anonitored whenever the rniisul.uinn pumps are not operating.
- Top of the active fuel is defined to be 360 inches above vessel zero (see Bases 3.2).
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References
" Generic Reload fuel Applications," NEDE-24011-P-A*
y, ,
8 Approved revision number at time reload fuel analys'es are p'erforced.
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A QIf Al).CliIl3 U DPR-29 2.11.lMITING SAIT.1Y SYS1CM SETTING BASES The abnormal operation.it transienis arplicable to operation of the units have been analyred throughout the spettrum of planned operatiag (onditions up to the rated thermal power condition of 25 II MWI. In add MWt is the licensed masimum steady state power level of the units.This maximum steady state power level will never knowingly be exceeded.
Conservatism inncorporated in the transient analyses in estimating the controlling factors, such as void coeflicient tontrol rpd uram
- orth, scram delay time. peaking factors. and axial power shapes. These facto selected comervativelv with resriect to their effect on the applicable trarnient re analysis model.
is documented in Reference 1. Transient analyses are initiated at the conditions given in this Reference.
the absolute vJtue of the voto reJettvity coemetent useo in tne anaryns is conservatively estimateQ to oc anout a3',o greater than the nominal maximum value expected to occur during the core lifetime. The scram wort been derated to be equivalent to approximately 80" of the total scram worth of the control rods.1he scram delay time and rate of rod imertion allowed by the analyses and conservatively set equal to the longe >t delay and >lowe>t insertion rate acceptable by technical specifications. The effects of scram worth, scram delay time. and rod inse rate. all conservatn ely applied, are of greatest significance in the early portion of the negative reactivity insertion.
The rapid insertion of negaine reactivity is assured by the time requirements for 5% and 20%insertson. By time the rods are 607 mserted, approximately 4 dollars of r egative reactivity have been inserted. which strongly s suins the transient and accomplishes the desired effect. The times for 50% and 90% insertion are given to assure proper completion of the expected performance in the earlier portion of the transient. and to estahtish fully shut down steady 4 tate condition.
This choice of using conservative values of controlling parameters and initiating tiansients at the design power level proJuees more pessimatic anwers than would result by using expected values of control parameters and analyzing at higher power levels.
u Steady. state operation w ithout forced recirculation will not be permitted except during start' p testing.The ana 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 anal 3ses 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 e)c'ist prior to initiation of the transients.
A. Neutron I'lus Trip Settings
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 fission chambers preside the basic input signals. the APRM system responds directly to average neutron flus. During transients. the mttantanecius rate of heat transfer from the fuel (reactor thermal powei) is less than the instantaneous neutron flux due to the time constant of the fuel. Therefore. during abnormst t perational transients, the thermal power of the fuel will be less than that indicated by the neutron Ihn at the u ram sening Analyses demonstrate that with a 120% scram trip setting. none of the abnormal oper.monal tr.insients analyred violates the fuel safety hmit. and there is a substantial margin f rom fuel damage.Therefore. the use of flow referenced scram trip provides even additional
.I mar gin.
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QUAD-CITIES
- l DPR-29 -- - l
fuel cleJJang integrity natety limit, is reached. The APRM scram trip setting was determined by en analysis of margins required to provide a reasonabic range for maneuvering during operation. Reducing this operating margin would increase the frequency of spurious scrams,
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.V which have an adverse ef f ect on reactor safety becaum of the resulting ther :1 stresses.
Thus, the AP M meram trip setting was selected I;ocauce it provides adesuste margin for the reduces the posstbil-fuel cladJang integrtty safety limit yet allows operating n argin that ity of unnecessary scrams.
l The scram trip setting must be adjusted to ensure that the IJIGR transient peak is not increased for any combination of maximum f raction of limiting power density (MFI.PD) and I reactor core ther:ral powcr. The scram setting is adjusted in accordance with the formula in Specification 2.1. A.1, when the MFLPD is greater than the fraction of rated power (FR2).
- The adjustment may be accomplished by increasing the APRM qain by the reciprocal of FHP/MPLPD. Th i c; provides the same degree of protection as reducing the trip setting by FRP/MFLPD by raising the initial APRM readings closer to the trip settings such that a scram would be received at the same point in a transient as if the trip settings had been re-duced by Fal(__
MFLPD* ,
,, 2. APM Fluu 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 settir of 15% of rated power provides adequate thermal margin between the setpoint and the safety
! limit , 25'i of rated. The margin is adequate to acco=odate anticipated maneuvers associated with power plant startup. Effects of increasing pressure at zero or low V0id c ntent 3rc >
minor, col t ater f r: .. wri evni hbh during startup 1J not much colder 1,uan tu almdy in tN system, t em.*: t a t t. r.: c.c cita sents are small, and control rod patterns are constrained to be uniform by operating proco,Nres La:md up by the rod worth minimizer. Of all possible source ,
' of reactivity input, uniform control rod withdrawal is the most probable cause of significan:
l power rine. Because the flux dictribution associated with uniform red withdrawals does not i involve high local peaks, and because several rods must be moved to change power by a signif t
- cant percentage of rated power, the rate of power rise is very clow. Generally, the heat flu
- la in near equilibrium with the fission rate. In an assumed uniform rod withder.wal approach
- to the scram level, the rate of power rise is no more than 5% of rated power per minute, and the APRM system would be more than adequat'e 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 pressura is greater than 850 psig.
l 3, IRM Flux Scram Trip setting l
} The IM system consists of eight chambers, four in each of the reactor protection system log:
. channels. The IM is a 5-decado instrument which covers the range of power level between thz j covered by the SRM and the APRM. The 5 decades are broken down into 10 ranges, each being one-half a decade in size.
The IRM scram trip setting of 120 divisions is active in eacit range of the IRM. For example,
! if the instrunent were on Range 1, the scram setting would be 120 divisions for that range r I likewise, if the instrument were on Range 5, the scram would be 120 divisions on that range.
j Thus, as the IRM is ranged up to accommodate the increase in power level, the scram trip set-ting is also ranged up.
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The mest significant cources of reactivity chango during the power increase are due to contri rod witt.!rawl. In order to ensure that the IRM provides adeauate protection against the single rod wtthdrawal error, a range of rod withdrawal accidents was analyzed. This analysi:
included starting the accident at various power levels. The most severe case involves an t
initial condition in which the reactor is just subcritical and the IRM system is not yet on l I scale.
Additional connetvatin was taken in this analysis by anruming that the IRM chsenelClosest '
- 4. t he w t '.$. !r iwn t o t v. b; , med . The results of this malva ts show that the reactor is scrc=-
-g ,a nd p. m pa c t t am tt.*J to 1;. of rated power, thus mainta'ining MCPR above the fuel cladding w l inteqrtty mately lintt. pased on the above analysis, the IRM provides protection against local control ro<l withdrawal errors and continuous withdrawal of control rods in sequence an a provides backup protection for the APRM.
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. QUAD-CITIES
. . DDR-29
(;@y:
O S. APM Rod alock Trip setting Reactor power level may be varied by moving control rods or by varytng the recirculatien flow rate. The APM system provides a control rod block to prevent gr~ss rod withdrawal at constant
' recirculation flow rate to protect against grossly exceeding the MCPR ruel Cladding Integrity Safety Lusit. This rod block trip setting, which is automatically varied with recirculation loop flow rate, prevents an increase in the reactor powar level to excessive values due to control god withdrawal. The flew 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 relationshipr therefore the worst-case MCPR which could occur during steady-state operation is at 10e% of rated thermal power because of the APRM rod block trip
. setting. The actual power distrtbution in the core is established by specified control red sequences and is monitored continuously by the incore LPM system. As with APM scram trip '
setting, the APRM rod block trip setting is adjusted downward if the maximum fraction of limit-ing power density exceeds the fraction of rated power, thus preserving t.6.e 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 scra's setroint is based on normal operat-ing temperature and pressure conditions because the level instrt. mentation is density compensated.
, D. Reactor Low Low water Level ECCS Initiation Trip Point The emergency core cooling subsystems are designed to provsde sufficient cooling to the core to dissipate the energy associa ed withthe loss-of-coolant accident and to licit fuel claddin6 temperature to well below the cladding melting temperature to assure that core geometry remains
? intact and to ILmit any cladding eietal-water reaction to less than 1L To accomplish their
- .7; . 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 set point of the low water h level scram would increase the capacity reauirement for each of the ICCs 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 def endent on three previously a
set parameters: the maxi =cm break size, the low water level scram setpoint, and the ZCCS initiation setpoint. To lower the setpoin*. for initiation of the ECCS could lead to a loss of .
effective core cooling. To raise the ECCS initiation setpoint would be in a safe direction, but it would reduce the margin established to prevent actuation of the ECCS during nonnal
- operation or during normally expected transients.
- 3. durbineStopvalveScram i The turbine stop valve closure scram trip anticipates the pressure, neutron flux, and heat flux increase that could result from rapid closure of the turbine stop valves. With a scram trip setting of 10% of valve closure from full open, the resultant ancrease in surface heet flux is limited such that MCPR remains above the MCPP. fuel cladding integrity safety limit even duriag g the worst-case transient that assum-s the turbine bypass is closed.
- y. Turbine control valve Fast closure scram The turbine control valve fast closure scram is provided to anticipate the rapid increase in pressure and neutron flux resulting from f ast closure of the turbine control valves due to a load rejection and subsecuent failure of the bypass, i.e., it prevents MCPR from b ccming less than the MCPR fuel cladd.nq integrity safety limit for this transient. For the 1oad rejection without bypass transient from 100% power, the peak heat flux (and theiefore LHGR) increases on the order of 15% which provides wide margin to the value corresponding to 1% plastic strain of the cladding.
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G. Reacter Coolant Low Preware Initiates Maln' Steam Isolation Yahe Closure De low pressure ital.ition at 850 psig was provided tu give pr occurs in the Run mode when the main steamline isolation valves are closed to provide for reac shuidown so that oper. tion at pressures lower than those specified in the thermal hydra does not occur, ahhough operation at a pressure low than R50 psig would not necessarily c ensafe condition.
H. Main Steamline Isolaiian su Valse Closure Scram ,
The low pressure isnlJtion of the main steamlines d to at the scram feature in the Run mode which occurs when the main steam previde for reactor shutdown so that high power operation at low reactor press providing lower thn 1:
prutection for the fuel cladding integrity safety limit.
of the fuel staddmg integrity safety limit is provided by the IRM and APRM high neutron flux sc Thus. the combinatson of main steamline low-pre >sure isolation and isolation valve clos Run mode assures the availability of neutron flux scram protection over the entire range ,
of the fuel cladding integrity safety limit. in addition, the isolation valve closure scram in anticipates the pressure and flux transients which occur during no increase normal orl inadve in neutron einsure. With the scrams set at 10% valve closure in the Run mode, there is flux.
- 1. Turbine F.HC Control Fluid Imw-Pressure Scram The turbine EHC control system operates using high pressure oil. Dere are several poin system where a lost of oil pressure could result in a fast closure of the turbine co closure of the turbine control valves is not protected by the turbine control valve fast closur failure of the oil system would not r-sult in the fast closure solenoid valves being actuated. F control valve l'ast closure the core would he protected by the APRM and high reactor pressure scram However. to provide the same margins as pmvided for the generator load f rejecibn on f turbine control valves. a scram has been added to tne reactor protection system which sense control oil prenure to the turbine control system. This is an a to that resulting from the turbine control valve fast closure scram. The scram setpoint of 9 high enough to provide the necessary anticipatory function and low enough to m spurious scrams. Normal operating pressure for this system is 1250 psig. Finally not start until the fluid pressure is 600 p<ig. Therefore, the scram occurs well before valve clo bef in>.
J. Condenser Low Vacuum Scram Loss of condenser vacuum occurs when the condenser can no longer handle the hea cor> denser vacuum initiates a closure of the turbine stop valves and turbine bypass v eliminates the heat input to the enndenser. Closure of the turbbie stop and bypass f valves causes transient. neutron thex rise, and an increase in surface heat flux.To prevent the cladding s i being exceeded if this occurs a reactor scram occurs on turbine stop valve closure
, sutbine stop valve closure scram function alone is adequate to prevent the cladding ) safi l
bcing execeded in the event of a turbine trip transient with bypass closure. l The condemer luw vacuum scram is anticipatory to the stop valve closure scram and causes al
("
before the stop valves are closed and thus the sesuhing transient is less severe. Scram mode at 23 inctill f
, vacuum stop valve closure occurs at 20-inch Hg vacuum, and bypas Hg vacuum.
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I References -
- 1. " Generic Roload Fuel Application," 2TEDE-240ll-P-A* i
- Approved revision number at time reload analyses are performed i l
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QUAI)-Cllig'.S opa-?ct 1.2 SMI'.TY IJMll IIASIW ,j The reastor coolani sysicm mict,rity is an important barrict in the prne ion of unconticl!cd sc! case of fmion products 14 is cucntial that the inteprity of shis system be protetied by estabbshinp a pressure hmit to be observcd for a!! opera;ing conditions and wheneser there is irradiated fact in the reactor vessel.
De prenure safety limit of 1325 pig as rne.isurcJ by the vessel sicam space preuure indicator is equivalent to 1375 pig at the larst elevarian of the reattor coo! ant systesn lhe 1175 pig va:ue is detacd from the design preuurcs of the se.ittur pecuurc venel and coolant systern piping.1he ic>pective dengn preuures are 1250 rig e
at 575* l' and 18 75 pig at Se,0 I . lhe prenure saf(tv limit was chosen as .he lower of the preoure tr.tnsients permitted by the appheable design codes: ASMl; Boiler sud Pressure Venel Code Section 111 for the p<euure sc .cl.
and USASI1t31.1 C.xle for the reactor coolant system piping The ASMt.16ter and P suure Venel Code permits pressure transients up to In';.oser desi;n pressure (1I0". a im = 1375 psip and the 1:NA%I Code permits pressure transients up to 20% oscr the design pressure (120"o a 1I75 - 1410 psip) lhe safety limit pressure of 1375 psig is referenced to the lowest elevation of the primary coolant system. Evaluation methodology to assure that thic cafety limit pressure is not exceeded for any reload is documented in Reference 1.
The design basis for the reactor preuure vcwI niales evident the substanual mastin of proicction against failure at the safety pressure hmst of 1375 pig. lhe vessel has becn desivned for a reneral membrane stren no prcater than 26.700 psi at an internal preuure of 1250 psig, this is a factar of f.S below th< yield strenph of 40.100 psi at $75
- It At the pressure limit of I 375 psig, the general membrane stren will only be 29.400 pi. Sisti safely below the yield strength.
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The relationship of stress leveh to yield strength are comparahle for the prirnary system pipin; and provide r similar margm of protection at the establis!>cd safcty presme hmit. ,
F The nonnel opcratmp pressure of the scactor coolant system is 1030 ps4. Ior the tuitm.: tilp orloss of electiicalload
' transients, the turbine trip scram os pnerator load rejection suam tor, ether with the turbi.te bypass sysicm limits the pressuse to approttmately 1100 pua. (itefciences2,3 and4) In addition. rressute rel<f valves have been rio'ided to reduce the probabdity of the safety ahes operatuig m the event that the tuibine bypass sheuld fad.
Feially, the safety valves are sued to keep the reactor coolant systs'm preuu
- tetow 1375 psig uth no credit taten for relef valves dunns the postulated full closure of all MSIVs wahout duett (salu position switch) scram. Credit is taken lov the neutron flus scism, however. l The indirect flux scram and safety vahe actuation, provide adequate margin i below the peak allow able veswl pressure of I 375 psg.
Reactor pressume is continuously monitored in the control room during operation on a 1500 pi full.seate pressure recorder.
Refe ences 6.
- l. " Generic Heload Fuel Application", NCDU-240ll-P-A*
- 2. SAR, Section 11.22
- 3. Ound Citics 1 Nttelcar Power Station firr.t reload licenne , '
submittal, Section 6.2.4.2, February 1974.
4 GE Topical Itoport NEDO-20693, General Electric Dolling Water Reactor No.1 licencing submittal for Quad Citics Nucicar Power Station Unit 2, December 1974 '
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- Approved revision number at timo relodd ianalyr,co are performed. '
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, QUAD CITIES DPR- 29 P.2 LIMITING SAFETY SYSTEM SETTING BASES In compliance with Sectlan III of the ASME Code, the safety' valves must be set to open at no hi' c.ber than 1037, or design pressure, and they must limit the reac tor pressure t a,no . fore than ))O"4 of design pressure. !!oth the high neutrcn flux scram and safety vn)ve actuation are required to prevent overpressurizing the reac:.or pressure vessel and thus exceeding the pressure rarety limit. The pressure scram is available as backup protect!on to the high flux scram. Analyses are performed as described in the
" Generic Reload Fuel Application," NEDE-240ll-P-A (approved revision number at time reload analyses are performed) for each reload to assure i
- -tKat the pressure safety limit is not er.ceeded. If the high-flux scra.i were to fail, a high-pressure scram would occur at 1060 psig.
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- 3.1/4.1 IIEACTOlt P!tOTECTION SYSTEM ljMlllNG CONIllTIONS FOR OPERATION SURVEll.l.ANCE RFQti!RDIENTS I.
Applicabiht):
- Applicaliility:
" Apphes to the surveillance et the instrumentation Applie, to the snurumentat:on anJ assretiated d.-
uces which initiate a reactor scram.
and amwiatcJ devices which initiate reactor scram.
Objecthe: Objecthe:
To auure the operability of the reactm pro'eetion To specify the type and frequency nf surveillance to
- - be applied in the protection inurumentauon.
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- SPECIFICATIONS
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t The setpoints. ndnimum number of trip sys. A. Instrumentation systems shall be functior. ally tems, and minimum number of instrument tested and calibrated as indicated in Tables
- 4 l 1.md 4.1-2 respectively.
channels that must be operable for each posi.
A tion of the reactnr mode switch shall he as
- [v;) -
4 V.# given in Tahtes 3.1 1 through 3.1 4.The system B. D.dly during reacto: powcr operation. the core
! pow er distributiun shall be checked l'or ttaximum
- respon<e times from the opening of the sensor
- contact up to and includin; the opening of the fraction of limiting power dens-trip actuator contacts shall nnt exceed 'i0 ity (MFLPD) and compared with the
{ g. 8" " fraction of rated power (FRP) 4 l B* If, during operation, the maximum when operating ai>ove 25% rated fraction of limiting power dens- thermal power. _ . . . . .
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" power when operating above 25% C.
\Vhen a.t is de:esmined that a channel is failed rated thermal power, either: in the unsafe cr'nJition and Colume 1 ot sa.
.l 1. the APRM scram and rod bles 3.1-1 through 3.13 cannot be net, that i; '
trip system must be put in the inpped cindition Ii - bloc)c settings chall be
~ immediatcly. All other RPS channch that man-1.* reduced to the values stor the same vanable sha'.I i:e (.n(ti, ai:v i{ given by the cc:uations teced witidn R h.,u:s The tnn v.st. . wuh S.
s in Sp::cificationu 2 . l . .'s .1 fai;cd ih innel may be untripp:d for a pened Of l '* and 2.1.D. This r.vty alSO time not to ciceed I hour to conduct this, be acc'ompilShed by testing. As Ions as the trip system with the failed channel contains at least one operahie increasing the APIU4 channel monitoring that same variable. that
' gain as dQSCribCd. l therein. IP. 5)5'em m 57 he P accJ in the unt'iPred position for short periods of time to allow functional aesting ur all RPS instrument chan.
nc!s as specified by Tat:e 4.1.I.The trip system
, ,- may be in th untripped position fnr no n nre than S hours per functior.at tot pent I for this
'*"I"O n 2. ths- power di.tr.ibution l
() shall bec chang.d Lucit that the maximu:n fraction of limitinct power den::ity j
no lonUcr exct:cdu the l ,
fraction or rated power.
- 3.1/4.1-1 _
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QUAD-CITIES DPR-29 gallons. As indicated above, there is sufficient volume in the piping to accommodate the scram wit of the scram times or -nount ofinsertion of the control rods. This function shuts the react volume 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 initiates a closure of the turb;ne stop valves and turbine bypass valves. which eliminates the heat inpu condenser. Closure of the turbine stop and byp:ss valves causes a pressure transient. neutron flux rise increase in surface heat flux.To prevent the cladding safety limit from being exceeded if thh occurs, a reactor sc occurs on turbine stop valve closure. The turbine stop valve closure scram function alone is adequate to prev the cladding safety limit from being exceeded in the event of a turbine trip tr scram before the stop valves are closed, thus the resulting transient is less severe. Scram occurs a vacuum, stop valve closure occurs at 20 inches Hg vacuum. and bypass closure at 7 inches Hg vacuuta High radiation levels in the main steamline tunnel above that due background.The purpose of this scram is to reduce the source of such radiation to the extent r.ec excessive turbine contamination. Discharge of excessive amounts of radioactivity to the site environs i by the air ejector off. gas monitors which cause an isolatio t of the main condenser off. gas specified in Specification 3.8 is exceeded.
The main steamline isolation valve closure scram is set to scram when the isolation full open.This scram anticipates the pressure and flux transient which would occur when the val scramming at this settinf. the resultant transient is insignificant.
A reactor mode switch is provided which actuates or hypasses the various scram functions appropriate c O" particut.ir plant operating status (reference SAR Section 7.7.1.2). Whenever the reactor mode Refuel or Startup/ Hot Standby position, the turbine condenser low vacuum scram i and main steamline valve closure scram are bypassed.This bypass has been provided for flexibility during startup flux and to allow repa rs to be made to the turbine condenser. While this bypass is in effect protection is provided against pressure or increases by the high-pressure scram and APRM 15% scram, respectively, which are ciTective in this mo If the reactor were brought to a hot standby condition for repairs to the turbine condenser, the main steam isolation valves would be closed. No hypothesized single failure or single operator action in this mode of o can result in an unreviewed radiological release.
The manual scram function is active in all modes, thus providing for a manual means of tapidly inserting co rods during all modes of reactor operation.
The IRM system provides protection against excessive power levels and short reator periods in the intermediate power ranges (reference SAR Sections 7.4.4.2 and 7.4.4.3). A source i range monitor (SRM is aho provided to supply additional neutron level information during startup dbutinhas no scram funct (reference SAR Section 7.4].2 L Thus the IRM is required in the Refuel and Startup/ Hot Standby mo et addition. protection is provided in this range by the APRM 15% scram as discussed in the bases f 2.l. In the power range.the APRM system provides required protection (reference SAR Section 7.4 IRM system is not required in the Run mode. the APRM 's coveronly the intermediate and power r provWe adequate coverare in the startup and inte mediate range.
I The hi;h reactor preuure, high-drywell pressure, reactor low water level, and scram discharge vo scrams are required for the Startup/l ot Standby and Run modes of plant operation.They are therefor to be operational for these modes of reactor operation.
The turbine condenser low-vacuum scram is required only during power operation and must be hypassed to start
.;. the unit.
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41 SURVEll. LANCE REQUIREMENTS BASES -
A. The minimum functional testing frequency used in this trecification is based on a reliability analysis using the concepts developed in Reference 1.This concept was specifically adapted to the one-out.
taken twice logie of the reactor protection system The analysis shows that the sensors are primarity responsib!c for the reliability of the reactor protection system.This analysis mases use of* unsafe
- rate experience at conventional and nuclear power plants in a reliability model for the system. An e
failure' is defined as one which negates ch.inneloperability and which,du' to its nature,is revealed only when the channelis functionally tested or attempts in respond to a ter! signal. Failures such as blown
, fuses, ruptured bourdon tubes. faulted amplifiers, faulted cables. etc., which result in ' upscale
- or
'downseale* readings on the reactor instrumentation are 'safc* and will be casily recognized by the operators during operation heeause they are revealed by an alarm or a scram.
The channels listed in Tables 4.1-l and 4.12 are divided into three groups respecting functional testin These are:
I. on.off sensors that provide a scram trip function (Group 1 h
- 2. analog devices coupled with bistable trips that provide a scram function (Group 2 g and
- 3. devices which serve a useful function only during some restricted mode of opera: ion. such as Startup/ Hot St.indby. Refuel, or Shutdown, or for which the only poetical test is one that can be .
performed at shutdawn (Group 3). ,
The sensors that mal 6 up Group I are specifically selected from among the whole family ofindustrial on-off sensors that have earned an excellent reputation for reliah*e operation. Actual history on this class
-G of sensors operating in nLclear power plants shows four failures in 472 sensor years, or a failure rat V
0.97 x 10*/hr. During design,a goal of 0.99999 probability of success (at the 50% confidence level) was adopted to suure that a balanced and adequaic design is achieved.The probability of success i a function of the senton failure rate .ind the test irnerval. A 3-month test interval was planned for Group I sensors.Thi> .s in Leeping with goad operating practee and satisfies the design goal for the logic configuration utilized in the reactor protection system.
To satisfy the long-tct n objective of maintaining an adequate level of safety throughout the plant lifetime, a minimum goal oro.9999 at the 95% confidence levelis proposed. With the one-out of-two taken twice logic, this requires that e.ich sensor have an availability of 0.993 at the 95% confidence le This level of availability may he maintained by adjusting the test intervsl as a function of the observed failure history (Reference 11.To facilitate the imp!cmentation of this technique. Figure 4.1 1 isl pro to indicate en appropriate trend in test interval. The procedure is as follows:
- 1. Like sensors are pooled into one group for the purpose of data arquisition.
- 2. The factor M is the exposure hours and is equal to the number cf sensors in a group, n, times the elapsed time T(M = nT).
- 3. The accumulated number of unsafe fai'ures 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 will be the test interval to the left of the plotted points. ~
- 5. A test interval of I month will be used ini;ially until a trend is established.
Group 2 devices utilize an analog sensor followed by an amplifier and a histable trip circuit.The sen and amplifier are active components, and a failure is almost always accompanied by an alarm indication of the source of trouble. In the event of failure, repair or substitution can start immediately.
l h An 'as.is' failure is one that ' Sticks
- midscale and is not capable of going either up or down in response 3.1/ 4.1 -5 l
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I QUAD-CITIF.S -
DPR-29 O switches. bence calibration is not upplicable i.e the switch is either on or off. Based on the shove, no calibration is required for thew instru. ment thannek. ,
l B. The MFLPD shall be cheded once per day to determine if the APRM scram requires adjustment. This may normally be done by checking the LPRM readings. TIP traces, or process computer calculations. Only a small number of control rods arc moved daily. thus the peaking factors are not expected to change MFLPD is adequate. l signific.intly and a daily check of the References I. I. M. Jacobs.' Reliability of Engineered Safety Features as a Function of Testing Frequency.' Nuc/ car Vol. 9, No. 4, pp. 310 312. July. August 1968.
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- . 3.2 LIMITING CONDITIONS FOR OPERATION BASES In addition to reactor protection instrumentation which initiates a reactor scram. protective instrument.ition has been provided which iniaates action to mitigate the consequences of accidents which are beyond the operator's abihty to control, or terminates operator errors before they result in serious consequences. This set of specifications prcvides the limiting condaisons of operation for the pnmary system isolation function. initiation of the emercency core coohng system control red blod. and standby pas treatment systemt The objectives of the specifications are (I) to assure the eflecineness of the protective instrumentation when required by preserving its capabihty to tolerate a single failure of any component of such spicms even during periods w hen portions of such 9 stems are out of service for mamtenance. and (2 ) to prescribe the inp sett n,es required to assure adequate performance.
When necessary, one channel may be made inoperable for briefintervals to conduct required funcuonal tests snd calibrations. Some of the settings on the instrumentation that initiates or controls core and containment cochng have tolerances explicitly stated where the high and low values are both critical and may have a substantial cifect on safety. It should be noted that the setpoints of other instrumentation. where only the high or low end of the setting has a direct bearing on safety. are chosen at a level away from the normal operating range to prevent inadvertent actuation of the safety system involved and exposure to abnormal situations.
Isolation valves are installed in those lines that penetrate the primary containment and must be isolated durinc a loss-of-coolant .ccident so that the radiation dose hmits are not exceeded durinc an accident condition. Actuation of these valves is initiated by the protective instrumentation which senses tre conditions for which nolatten is required ( this instrumentation is show n in Table 3.2-1 ). Such instrumentation rnust be available whenever pnmary containment integrity is required. The objectne is to isolate the primary containment so that the guidehnes of ill Cl-R IM are not exceeded during an accident.
The instrumentation which initiates primary system isolation is connected in a dual bus arrangement. Thus the docuwion given m the bases for Specification 3.1 is applicable here.
'Ihe low-reactor water level instrumentation is set to trip at 28 inches on the level instrtrnent (top of active fuel is defined to te 360 inches above vensel zero) and af ter allowinq for the full power prennure drop l across the steam dryer the icw level trip is at 504 inchen above v"scel 7ero, or 144 inches atuve top of active fuel, tsetrofit Ax8 fuel has an active fuel lenqth 1.24 inches Imier than earlier fuel designs, hrwever, present trip setpoints were usal in the TDCA analysis.* This trip initiatec clocure of nroup 2 and 1 litimry contain- l rnent icolation valven but (kos not trip the twirculation pumiv (reference FM, Smtion 7.7.2) . Ibr a trip setting of 504 inchen dove vesnel zero and a 60-ccconi valve clocure time, the valven will te closni lcfere perferation of the claddity occurn even for the Eximum break. "he cet t irn is, t her efore, a lo pate.
The low low reactor level instrumentation is set to trip when reactor water level is 444 inches above vessel zero (with top of active fuel defined as 360 inches above vessel zero,-59"is M inches above the top of active fuel).
This trip initiates closure of Group 1 primary containment isolation valves (reference SAR Section 7.7.2.2) and also activates the ECC subsystems, starts the emercency diesel generator, and trips the recirculation pumps.
This trip setting level was chosen to be high enougn to prevent spurious operation but low enough to initiate ECCS operation and primary system isolation no that no melting of the fuel cladding will occur and so that postaccident cooling can be accomplished and the cuidelines of 10 CFR 100 will not be exceeded. For the com-plete circumferential break of a 28. inch recirculation line and with the trip setting given above.ECCS mitation and primary system inolation are initiated and in time so meet the above criteria.
.The instrumentation also covers the full spectrum of breaks and meets the above criteria.
.' e i
r Locs of coolant necident analysis for Dresden Unit 2/3 & Quad Cities Units 1/2,
', NEDCL2hlhCA April,1979 t
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Vemuri tubes are provided in the main steamlinesi as a rueans nitoring steam ofDow.
measuring ste ;
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of mass inventory ne from the sewel during a steamline hreJL acc main sie.imline. thus only Gruup 1 valves are closed. For the worst 4ase l
j How, in conjunction instamemation is to detect a break !
acrident. main steamline breat nutside the drywell this i trip setting loss such ofthat120%
fuel is of notrated steam with the Dow limiters and main steamhne valse closure, limits the mass h i nventory is well beloa uncovered. fuel temperatures remain less than 1500* F. and release of radioactivity to 10 CFR 100 Cuidelines (refe: ente SAR Sections 14.2.3.9 andi 14.2.3.10).
Temperature. monitoring instrumentation is provideil in the l main I
eteaml Trips are provided on this instrumentation and when circeded cause clo ble ofcovering the entire setting of 200' F is low enough to detect leaks of the order of above.
5 to 10 and gpm; thus it is i disemsed for idelines of ID Cl R 100 spectrum of breaks. For large breaks. it is a b.ickup to high.
are exceeded. This High. radiation monitors in the main steamline tunnel have been provid ident. With the instrumentation causes closure of Group i valves, the only l valves required fusion product to close f SAR Section estJhlished setting of 7 times normal bEk$round and mJin 12.2.1.71 4 i f Pressure instrumentation is provided which trips when main steamline preuu this instrumentation resuhs in closure of Group i isolation valveti against in the a pressure regulimr this trip function is hypassed This function is provided primarily to 850 provik protect on psig. imeniory malfunction which would cause the control and/or hypass valve to open.h With the trip s loss is limited so that fuelis not uncovered and peal cladding treference SAR temperatures Section are are no fission products available for release other than those in the reactor water (V) I l .2.3 ). hi The RCIC and the llPCI high flow and temperature instrumentation IIPCI isolation vatves are provi respective piping. Tripping of this instrumentation results in actuation tf the RCIC Tripping logic for this function is the same as th.ii for th. main f2WI Fsteamline and iso required to be operable ur m a inpped c:mdition to meet t is within limits.
The instrumentation which initiates ECCS action is arranged h function rather than the in two a tripone.out o the reactor scram circuits, however, there is one trip system associated with h deac d ntcore systems in the reactor protection system.The i single. failure criteria are met toohng functions are provided. e.g., sprays and automat c declared inoperable.
he specification requires that if a trip system h tf henew of the becomes 53 stem ino out.of6ervice specifications of Specification 3.5 govern.Th3 specihcation pre >enes t e e et witti re>pect to the single radure criteria even during pernsds when maintenance or The control rod block functions are provided to prevent excessive control rod go below the MCPR Fuel Claddin,u Integrity Safety Limit.Th four SRM's will result in a rod block. The minimum h instrumentation to assure that the single. failure crueru are met. The miniraum inntrument c a instruntent el requirements for the RBM may be reduced by one for a short period of time to allow for maime This time period is only-3% of the operating time in a month and does not si preventing an inadvertent control rod withdrawal.
j O
3.2/4.26 f
l
). .. -
QUAD-CITIES DPR-29 so that none of'the activity released during th'e 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 3.2-4. 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 regardmg postaccident recovery.
The specifications allow for postaccident instrumentation to be out of service for a period of 7 days. His period is based on the fact that several diverse instruments are available for guiding the operator should an accident occur, on the low probability of an instrument being out of service and an accident occurring in the 7-day period, and cn engineering judgment.
The normal supply of air for the control rcor.1 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,'s.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. .
!. O t
l
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O 3.2 / 4.2 -8
QUAD-CITIES DPR-29 Q
Q The APRM rod block function is flow binced and prevent.r: a cignifiennt reduction in t'CPR, ce.pecially during operation nt reduced flow. The APRM provides gross core protection, i.e., limits the groos of control rods in the normal withdrawal sequence.
In the refuel and startup/ hot standby niodes, the APRM rod block function is set at 12% of ratec' power. This control rod block provides the same type of protection in the Refuel and Startup/ Hot Standby modes as the APRM flow-biar.ed rod block docc in the Run mode, i.e. , prevents control .
rod withdrawal bef ore a sc' ram is reached. . !
The RBM rod block function provides local protection of the core, i.e.,
the prevention of transition boiling in a local region of the core for a single rod siithdrawal error from a limiting control rod pattern. The trip point is flow biased. The worst-case single control rod withdrawal l
error is analyzed for each reload to assure that, with the spercific trip I
- settings, rod withdrawal is blocked before the MCPR reaches the fuel cladding integrity safety limit.
Below 30% power, the worst-case withdrawal of a single control rod with-out rod block ection will not violate tho fuol cladding interiri ty safety limit. Thus the RDM rod block function is not. required below this power level.
The IRM block function provides local as well as gross core protection.
The scaling arrangement is such that the trip setting is less than a factor of 10 above the indicated level. Analysis of the worst-case accident results in rod block action before MCPR approaches the MCPR fuel '
cladding intecarity safety limit.
J
- A dow nscale indi a
- inn on an APRM or IRM is an indie.ui m the instrument has faded or is r.ot semitive enaugh.
g In either case the inurument m e!I not respond to chanres in control rod motion, and the control rod motion is thus prevented The doe nscale inps are set 4: 3/125 of full seale. -
The SRM rod block with r 103 CPSand the detector not fully inserted assures that the SRM's are not withdrawn from the cose poor to emnmenomt rod with ! ramal foi stasiup The stram discharge vulume high w tcr kvel rod I bk,:L provides annunesasion for orciator action lhe alarm setpuint has been selu tcd to pionde adequate time l to allow cetermination of the cause s,flevel increase and conective action prior to autumane scrarn initiation.
- l. For errective et iergency core conhng for small pipe breats, the llPCI systern must function, since scaetor pressure l , does not decrease rapidly enough in allow ciiher core spray or I.PCI to operr.e in tune. lhe automatic pressure rehef function is provided as a bad up to the llPCI in ths e vent the llPCI does not opera.c 1he airang;ement of
} she tripping coni,en is suth as so pnwide this fuusNn w ben ncrewary and rmmmur sps rious operatmn 1he slip l
settings given in the specirwation air adequate to.- .ute the above enteria are met (rcicis ne c SAR sies non (. 2 A 3 )
1he speoficanon paesenes the eficsiiseness of the spicm afur.ng penods of mainten.n.sc. noting or uhbration and also mmimires the risk ofinadvertent operathn. e c.. only one instrument channel out of wirire.
Two air ejutor olli as t nmninns aic gnovidesi anil, when their trip point is reached. cau c un set.uion of the air I
- tjector on-pas hne intatmn n untiaird mhrn hnik imuumen% reash their higliinp pomt or our h.is an tipscale irip and the other a downwale inp 1 here is a l $.nunute sfelay hetoic the air ejector on-ras tvilainen valve sulosed.
This delay is accounied for 1.y the .Eniinute he.t.lup niin of ths off. gas ber>re it is rckased so the chienney.
Iloth inurumer.ts are requned for trip, hun the inusuments are so designed that any imirument f.nture gives a downseate inp lhe inp scuints of the mstrun.snas arc set so that the chimney rekase rate lun.t pven in Specif. cation 3 8 A 2 is not enseeJcd l'our radiation monitors are prusided in the reatior builJmg sentitation ducts whkh initiate iwlation of the rear *or budJmt and optration of the standh? ras tecatment spiem. lhe monitors are Imated in the reactor buiktmg senntatum sh.se 1he esip 1.Ti e is.i mn one of awo for cash sci. and cath set (Jn inilianc 3 Hip mdepenilent of the other % Any up s.de nip = di sause the siemed acomm Tnp seitmf.s of 7 enR/hr for nu.mton in the venidasion Jm t air b.no d i ,nn moi um noinul sconbnon not neon and uandby pn encannrns sysicm operatmn Q so than the gennlanon se n 6 sch se sate huni pren m *qwikatom 3 Ts A 3 is not essccdesi 1mo rastsatum momsors \
Q are providcJ on the n eus hne it.=n minst imtute i olai.on el the reassor bushtmr and "Priasmn of inc stanJhy gas tiraiment spt(un ihr u p l.in n ont out ut two ' hip wunp of 100 mR/In for the m-muun on the refuehnt ilm.: asc becJ up.m imo.omt n.nmal ss nid.aum notanon .md namthy ras nealms ut syurm recration 12/.L17
QUAD-CmES DPR-29 O -
TABLE 3J4 , .
MSTRUMENTATION TNAT INITIATES ROO BLOCK Malmun ihunter of Operatie er Trtsped lastrument .
Casasets por ing $rstem'" lastnesset Irtp Level Setting APRM inscale (flow besyn $[0.650W 2
D + 431 FRP MFLPD 2 APRM ipscale (Refuel at Statup/Het s12/125 fd scale Standby rnode) 2 APRM hwnscale'n 23/125 fui scale 1 Roe shck montts upscale (flow basP 50.550W + 42m t
1 Red bbck monitor dowitscale" 23/125 ful scant 3 RM downscale *
- 23/125 tti scale 3 m upscale * $108/125 fd scale 28 SW detector not si Startup positerf# 22 feet below core center.
he 3 m detector not n Statup position
- 22 teet bebw core center-he 28
- 3RM upscale s 105 counts /sec 2* 3RM downscate* 2 108 counts /sec 1 Hgh water level in scam docharge volume 525galons
, Notes
, 1. For the Startup/ Hot Standby and Hun positions of the reactor modo ec1ceter
' switch, there shall be two opc able or tripped trip systens fer cac:a fune-l tion except the SRK rod blochu. JIU4 upscale and IRM downscale need not be operable in the hun pMition, AttM down=cale, APRM upscale (flow biased),
and RnM downceale need not be opcrab]c in the Startup/ Hot Standuy mete. I i The RTE! upscale raced : ot be operabic ut less than 30% rated thermal power.
One chsnnel may be byrsr.ned above 30% rated thermal power provided tha t a limiting control roo pattern does not exist. For systems with more than I one channel per trip syst.em, if the first column cannot be met for one of I
the two trip systems, this condition may exist for up to 7 days provided that dtirinr, that tiene the operable' sy: tem is functionally tested in-mediately and daily thereorter; if this condition lastu longer than */ days the cystem shall be tripped. If the first column cannot be met for both trip systems, the systems shall be tripped.
- 2. W-d is the percent of drive flow required to produce a rated core flow of 9e million Ib/hr. Trip level setting is in percent of rated power (2511 Mdt).
1 mal demonsah mer he Woensed eben a e en c w react
& h tasten a messies ehes t% const sate e 2100 CPS.
1 Ibe ed me hw pas oe ts me,to typseest
& The 998 knctes may to bypassed a the higher IRM tentes (sanges 3.1 and W s%n the Sid specale red bieck h operable
- 7. not regned to be apewe .Me patermet Is= some phrus tests at atmosphes steenere dureg er ans Wwieg at sews levets not to escoed S Mwt.
1 Ibe ed haicten estes enen the reacts mode seitcti es a t% Reisela staitus/Not stendtry positen.
t . .. . d . . , _ ,ed 3.2/42-14
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QUAD-CITIES lil'P-29 O 1ABLE 3J4 .
P95TACC10ENT MONITORING lilSTRUMENTAhl0E REQUIREMENTS'il hstranset theisem aceter Reedest of Operable .
Lacetten Bomber taeneetin un parameter seit 1 Provtled Range i nesctor premure 901-5 1 S1500 piin 2 o.1200pse e nesctor ..tu wi 901-3 2 100 inches + 200 ince=s to inches is top of tuell
- 1 Terve etw tempw.ture 901-?1 2 0200* F 1 Terve oir temowsture 901-21 2 4600* F Terve wetc ieve'. 901-? 1 25 incfws - + 25 inches indicator
. 2 t43 Terve weier level. 1 14 inch range eight 94ss 1 Terve preswre 901-3 1 4 inene He to 5 pii, 1 Orywell preseste 901-1 1 4 nchee He to 5 psis O to 75 pig 2 Drywoti temperature 901-21 6 0400* F
! 2 peowtron monitoring 901-5 4 o.t.ia* ces 2"3 Terus to drywell 2 03 esid differentist preseare j
i i
1 Ideems
- 1. lastrument channels reauired during power coeretion to monitor pastaccident conditions.
l 2. Prorsions are mese for local sempling and monitoring of drywell etmosphere.
1 i
- Top of active fuel is defined to be 360 inches above vessel zero (See Bases 3.2). ,
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3.2/4.2-15
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I i
f QUAD-CmES DPR-29 O "
- 3. De control rod drive housing support 3. The correctness of the control rod system shall be in place during reactor withdrawal sequence input to the power operation and when the reactor RWM computer shall be verified after coolant system is pressurized above loading the sequence.
atmospheric pressure with fuel in the reactor vessel unless all control rods Prior to the start of control rod with-are fully mserted and Specification drawal towards criticality, the capabil-3.3.A.I is met. ity of the rod worth minimizer to properly fulfill its function shall be
- a. Control rod withdrawal sequences verified by the following checks:
shall be established so that max.
imum reactivity that could be a. The RWM computer online diag- '
~
added by dropout of any incre- nostic test shall be successfully ment of any one control blade performed.
would be such that the rod drop accident b. Proper annunciation of the selee-design limit of 280 cal /crm._is_not exceeded. tion error of one out-of-sequence
- b. Whenever the reactor is in the *"*"" **"
- Startup/ Hot Standby or Run c. The rod block function of the mode below 20% rated thermal 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 quali6ed technical person than to the block point.
may be used as a substitute for an inoperable rod worth minimizer which fails after withdrawal 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 onless at least startup or during refueling, verify that two source range channcis have an at least two source raoge 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 trol rod patterns, as determined by the exists, an instrument functional test of nuclear engineer. cither: the RBM shall be performed prior to
- b. control rod withdrawal shall be blocked; or O
3.3/43-3
, emk "9*
QUAD-CITIF.S firH-29 O c. the operating power level shall be limited so that the MCPR will re - (
j main above the MCPR fuel cladding integrity safety limit assuming a sin-gle error that results in complete withdrawal or;;ny single operable control rod.
- i C. Scram Insertion Times C. Serem insertion Times
- 1. The average wram inwrtion time. ha- 1. After refueling outage and prior to sed on the deenergiution of the scram operation above 307 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 i 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 Scram control rod drive pumps.
% imerted from insertion fully withdrawn Times (sec) 5 0.375 20 0.900 50 2.00 90 3.50 '
, O The average of the scram insertion j 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:
i
% inserted from Average Scram Ekelly Withdrawn Times (sec) 5 0.39N 20 0.954 50 2.12 90 3.80
- 2. The maximum scram insertion time 2. Following a controlled shutdown of for 90% insertion of any operable con- the reactor, but not more frequently trol rods shall not estved 7 seconds. than 16 weeks nor less frequently than 32 week intervals. 50% of the control
- 3. If Specification 3.3.C.I cannot be rnet, rod drives in each quadrant of the
! the reactor shall not be made super- t sMIW d for N critical; if cperating. the reactor shall ; ;pg g 3 ,
be shut down i,mmediately upon deter- 3.3.C. All control rod drives shall have mination that uverage scram time is ,F,;g ,, , 3 deficient. ccch year. Whenever all of the control
- 4. If Specification 33.C.2 sannot be met. tod drive scram times have been mea-the deficient control rod shall be con- sured, an c~aluation shall be made to q
D 3.3 /.t.3-4
- - - -s- -.
QUAD CITIES DPR-29 v
B. Control Rod Withdrawal
- 1. Control rod dropout acciderits as discussed in Reference 1 con lead to s16nificant core damacc. If coupling integrity is ma!ntoined, l the possibility of a rod dropout accident is climinated. The over-travel position feature provides a positive check, as only uncoupled drives may reach this position.
. N'utron instrumentation response to rod movement provides a verification th'at the rod is Miowing its drive. Absence of such response to drive movement would indicate an uncoupled condition.
- 2. The contro! rod housing support restricts the outward movement of a control rod to less than 3 inches in the cattemely rernote event of a housing failure. The amount of reactivity which could be added by this small amount of rod withdrawal, which is less than a normal single withdrawal increment. will not contribute to any damage to the primary coolant system lhe design basis is given in Section 6 61.and the design evaluation is given in Section 6.6 3 of the SAR. This support is not required if the reactor coolant system is at atrnospherie pressure, since there would then be no driving foret 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 demonstrated since the reactor would remain suberitical even in the event of complete ejection of the strongest control rod.
3 Control rod withdrawal and insertion sequences are established to ;
assure that the maximum insequence individual control rod or control '
rod segments which are withdrawn could not be worth cuough 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 sequences are developed prior to initial oper-ation of the urtit following any refueling outage and the requitement A that an operator follow thenc ricquences is stiperviced t>y the RWM or Q a second qualified ctation employce. These acquenecc are developed to limit ,rcactivity worthn of control rods and g
, l
> sogether with the integral rod velocity tirniters and the action of the control rod drive system.
Emits potential reactivity insertion such that the resolu ol a control rod drop accident wdl tiot esteed a maximum fuel energy content of 280 cal /gm lhe peal fuel enthalpy of 280 cal /tm as below t\r enert,y content at which rapid fuel siispersal and primary system damage have been found to occur based on esperimental data es is discussed in Reference 2 , I The analysis of the control rod drop accident was originally presented in Sections 7.9.3. H.2.1.1and 14.2.1.4 of the SAR. Imprmements in analytical capehihty have allowed a more refined analysis of the control roJ drop acuJent
,These techniques are described in a topical report (Reference 2) and l two supplements (References 3 and 4). In addition, a banked position withdrawal sequence descr$ bed in Reference 5 has been developed to further redur*c incremental rod worths. Method niid t>uuj u for the rod drop accident analyscc are documented in Reference 1.
l By using the analytical models dewribed in those reports soupled with conservative or worst. case input parameters, n has been determened that for power levels less th.inB% of rated power, the l specified limit on insequence control rod or control rod segment worths will limit the peak fuel enthalpy so less than 280 c.d/r. Above20% power even single oper.itor errors cannot result in g out<>f-sequence control rod moriht which are sufficient to reach a peat fuel enthalpy of 250 cal /g '
should a possulated control rod drop accident occur.
The following parameters and worst-case assumptions have been utilized peak fuel in the analysis to determine compliance with the 280 cal /gm l
enthalpy. Each core reload will be analyr.cd to show 1
conformance to the limiting parameters.
- a. an interassembly local peaking factor (Reference 6).
l 3 3/4.3-8
. QUAD CITIES IMH-2cl
- b. the delayed neutron fraction chosen for the bounding reactivity curve l
- c. a bestinning-of-life Doppler reactivity feedback
- d. scram times slower than the Technical Spec'ification rod scram insertion rate (Section 3 3.c.1) .
- e. the maximum possible rod drop velocity of 3 11 fps
- f. the design accident and scram reactivity shape function, and g the moderator temperature at which criticality occurs In most cases the worth of insequence rods or rod segments in con, function l with the actual values of the other important accident analysis parameters described ~above, would most likely result in a peak fuel enthalpy sub-stantially less than 280 cal /g design limit. ~l Should a control drop accident result in a peak fuel energy content cr280 cal /g. fewer than 660 (7 x *
- 7) fuel rods are conservatively estimated to perforate. This would result in an otsite dose we!I 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 diferences.
The rod worth minimizer provides automatic supervision to assure that out of sequence control rods will not be withdrawn or inserted; i.e it limits operator deviations from planned withdrawal sequences (reference SAR Section 7.9). It serves as a backup to procedural c:ntrol of control rod worth. In the event that the rod worth minimizer is out of service when regiaired, a licensed operator or other qualified icchnical employee can manually fulfill the control rod pattern conform.ince h
V function of the rod worth minimizer, in this case, the normal procedural controls are backed up by independent procedural controls to assure conformance.
A 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 emeient 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 10 8 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 contro! 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 centrol rod patterns, the withdrawal of a designated single control rod could result in one or more fuel rods wi th MCPh's less than the MCPR fuel cladding integrity safety limit.During use of such patterni it is judged that testing of the RBM system to assure its operability prior to withdrawal of such will assure that improper withdrawal does not occur. It 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 patierns.
3 3/4 3-9
QUAD CITIES OPR-29 l C. Scram Insertion Times
! The control rod system is analyzed to bring the reactor subcritical at
' a rate fast enough to prevent fuel damage, i.e., to prevent the MOPP.
from becoming less than the fuel cladding inte5rity safety limit.
I ^
i ~ Analysis of the' limiting power
~
transient shchs that the negative reactivity rates resulting from the scram with the average response of all the drives as given in the above specification, provide tne reouired protection, and MCPR re=ains greater than the fuel cladding inte5rity safety limit. The mimimum amount of reactivity to be inserted durirn a scram is controlled by permitting no trore than 109. of the cperable rods to have long scram times. In the analytical treatment of the transients, 290 milliseconds are l allowed between a neutron sensor reaching the scram noint and the start of r:nr ksn of the control rods. 'this is adequate and conservative when compared to the typically observed time delay of about 210 millisecorris. Approximately 90 l i milliseconds after neutron flux reaches the trip point, the pilot scram valve
' solenoid deenergizes and 120 milliseconds later the control rod motion is estimated to actually begin. tiowever, 200 milliseconds rather than 120 milliseconds is conservatively assumed for this time interval in the transient analyses and is also included in the allowable scram insertion times specified in Specification 3.3.C. The scram times for all control rods will be determined at the time of each refueling outage. A representative sample of control rods will be 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. The test schedule provides reasonable assurance of detection of slow drives before system detorioration I beyond limits of Specification.3.3.C. The program wan developed on the basis
- of the statistical approach outlined below and iudgment.
I - - . _ .
I i
i The history of drive performance accumulated to date indicates that the 90% insertion times of new and overhauled drives approximate a normal distribution about the mean which tends to become skewed soward longer scram times as operating time is accumulated.The probability of a drive not execeding the mean 90% insertion time by 0.75 seconds is greater than 0.999 for a normal distribution. The l measurement of the scram performance of the drives surrounding a drive exceeding the expected range of scram performance will detect local variatior.s and also provide assurance that local scram time limits are not exceeded. Continued monitoring of other drives exceeding the expected range of scram times provides surveillance of possibic anomalous performance.
'Ihe numerical values nstigned to the predicted scram performance are based on the analysis of the Dresden 2 startup data and of data from other ilWR's such as Nine Mile Point and Oyster Creek.
l
) The occurrence of scram times within the limits. but si tnifwantly longer than average, should be viewed l as an indication of a systematic problem with conuol rod drives. especially if the number of drives l exhibiting such wram times eseceds eight. the allowabic number ofinoperable rods.
l 3.3 / 4.3 -10 l
l i
- QUAD-CITIES
.* DPR-29 The occurrence of scram times within the limits, but significantly longer than average, should be viewed as an indication of a systematic problem with control rod drives, especia!!y if the number of drives exhibiting such scram times exceeds eight. the allowable number ofinoperable rods.
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 core calculations of a cold clean core. The worst case in a nine rod withdrawal sequence resulted in a k g < l.0. Other repeating rod sequences with more rods withdrawn resulted in k, > l.0. Ai reactor pressares in excess of 800 psig, even those control rods with inoperable accumulators will be able to meet required scram insertion times due to :he action of reactor pressure.
In addition, they may be normally inserted 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. Reactiilty Anomalies During each fuel cycle. excess operating reactivity varies as fuel depictes and as any burnable poison in supplementary controlis burned.The magnitude of this excess reactivity may be inferred from the critical tod 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 se nsitive and directly interpretable data relative to core reactivity. Furthermore, using power operating base conditions permits frequent re.ictivity comparisons.
Requiring a reactivity comparison at the specified frequency assures that a compatison will be made t
before the core reactivity change exceeds 1% ok. Deviations in core reactivity greater than IG .\k are I
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. .
i !
F. Economic Generation Control Speer,
! Operation of the facility with the economic generation control system (EGC)(automatic flow control)
{ is limited to the range of 65"o 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 of load 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 Rel,ad Fuel Applica tion", NEDE-24011-P-A" ,
- 2. C. J. Paone. R. C.Stirn.an'd J. A. Wooley.' Rod Drop Accident Analysis for L:rge BWRY GE Topical Report NLDO.10527. Marsh 1972.
. 3. C. J. Paone,il C.Stiro. and R. M Young.* Rod Drop Acciden: Analysis for Larre BWR*s'.Supp!cment 1.
OE Top!:al Repost Nr.DO.10527. July 1972.
f l. J. M. Havn. C. J. Paene, and R. C. Stirn. ' Rod Drop Accident Analysis for La'Ee BWR's, Addendum 2.
Esposed Coret.' Supp!cment 2, GE Topical Report NEDO 10527. January 1973.
- 5. c. J. Paone, " Banked position withdrawal sequence," 1,1 censing topical Report'UEDO-21231, January, 1977.
- 6. To include the power spike effect caused by gaps between fuel pellets.
. I
- Approved revinion runnber at t ime ecload fuel nrialyneu are perl'. nnel, e I j ,
- 3. w .1-ii j a ,
- QUAD CITIES DPR-29 q LIMITING CONDITIONS FOR OPERATION BASES b . 3.4 The design objective of the standby liquid control system is to A.
provide the capability of bringing the reactor from full power to a cold, xenon-free shutdown assuming that none of the withdrawn control rods can be inserted. To meet this objective, the liquid control system is designed to inject a quantity of boron which produces a concentration of no less than 600 ppm of boron in the j reactor core in approximately 90 to 120 minutes with imperfect mixing. A boron concentration of 600 ppm in the reactor core is required to bring the reactor from full power to 3% Ak or more )
subcritical condition considering the hot to cold reactivity swing, zenon poisoning and an additional margin in the reactor core for imperfect mixing of the I chemical solution in the reactor water. A normal quantity of 3470 gallons of solution having a 13.4"c 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 reactivity insertion due to cooldown of the reactor following the xenor. poison peak. I or a required pumping rate of 39 gpm, the maximum storage volume of the boron solution is established as 4875 gallons (195 gallons are contained below the pump suction and, thesefore, cannot be inserted).
Boron concentration, solution temperature, and volurne are checked on a frequency to assure a high reliability of operation of the system should it ever be required. Experience with pump operabthty indicates that monthly testing is adequate to detect if folures 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 system 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 Il V
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 inoperabic, there is no immediate ' treat to shutdown capability, and reactor operation may continue while repairs are being made. Assurance that the remaining system will perform its intended function and that the reliabihty of the system is good is obtained by demonstrating oper.:tmn of the pump in the operabic circuit at least once daily. A reliability analysis indicates that the plant can be operated safely in this manner for 7 dayt 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. The 10' F margin is included in Figure 3.3-1. 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 more water is added. Level indication and alarm indicate whether the solution volume has changed, wluch might indicate a posuble solution concentration change. Considering these factors, the test interval has been established.
O V
3.4/4.4-3 l
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, QUAll-CITIES DPR-29 0.
cycle by assuring that water can be run iluough the drain lines and actuating the air. operated vahes by operation of the following sensors: "
l) loss of air
- 2) equipment i sin sump high level
- 3) vault high level
- d. 'Ihe condenses pit 5.fint trip cir. .
cuits for each ch.mnet shall be checLed once a month. A lo;ic system functional test shall be per.
formed during each refueling ,
outage.
L Average Planar LiiGR L Ateente Planar LHCR During steady-state power operation, the average linear heat generation rate (APLHGR) of all the Daily during steady state operatir rods in any fuel assembly as a function of ave: age , aLbove 2 % rated thermal pcWer' p'anar exposure, at any axial location, shall not the average planar LHGR shall exceed the maximum average pianar LilGR be determined.
shown in Figure 3.51 If at any time l v during operathm it is determined by normal sur.
weillance that the firniting value for APLilGR is J. Local LHCR being exceeded, action shall be imtiated within 15 minutes to nestoic operation to .tithin the pre. Daily during steady state power operation sedbed lim.ts. If the APL11GR is not returned in above 25% of rated thermal power. the local 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 brought to the cold shutdown LHGR shall be determined.
condition withm 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Survedlance and corresponding action shall conttnue. untd reactor T operation is within the presenhed limits.
J. Local LHGR During steady state power operation, the linear heat generation rate (LilGR) of any rod in any fuel essembly at any axial location shall not '
exceed the maximum allowable LilGR.
If at any time during operation it is determined by normal survedlance that the hnuting vahac for LilGR is being exceeded, action sha!! be initiated wi.hin 15 minutes to restore operation to withm the pre-scribed limits. If the LilGR is not returned to i
)
l 3.5/4.59 l
l
. QUAD CITIES
. DPR-29 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 l
reactor shall be brought to the cold shutdown condition withm 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Surveillance and coe.
responding action _ sha!! contmue untd reactor -
operation is withm the prescribed limits.
Maximum allowable LHGR for all -
8x0 ruel types is 13 4 Kw/tt.
For 7X7 and mixed oxide fuel, the Maximum allowable LIIGR is as follows: .
j l LHGR o cLliG R, 1 -t aP/P),,,( L/L,)J where: - .-
LHG R, = , design LHGR
= 17.5 LW/ft.
( AP/P)_ = manimum power spiking penalty
" = .035 initial core fuel
= .029 reload I, 7 x 7 fuel .
I
= .028 reload 1,7 x 7 mixed oxide fuel L, = total core length
= 12 feet L = Axial distance from hottom of core K. Minimum Critical Power Ratio (MCPR) K. Minimum Critical Power Ratio (MCI'R)
During steady. state operation MCPR shall be The MCPR shall be determined d.uly durmf greater than or equal in steady 4 tate power operation .ihose 25% of 1.'5 (7 x 7 fuel) cated thermal power.
- 1.35 (8 x 8 fuel) at rated power and flow. If at any time during
. operation it is determmed by normal survedlance that the hmiting value for MCPR is being exceeded, action shall be initiated withm 15 minutes to l restore operation w within the prescribed limits.
If the steady. state MCPR is not returned to within the prescribed lunits 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 ,
i within the prescribed Inmits. For core flows other than rated, these nominal values of MCPR shall be increawd by a factos of kg where k g is as shown in Figure 3.5.2.
l
- 1 f
l 3.4/4.5-10 l .
l l
~
- QUAD. CITIES DPH-29 i
.'.5 a UhllTING CONDITION FOR OPERATION BASES .
A. Core Spray and LPCI alode of the RIiR S3 stem ,
his specification assures that adequate emergency cooling capability is available whenever irradiated fuel is in the reactor vessel. ._.. .. _
, Based on the loss.of<oolant analytical methods described in General Electric Topical Report*NEDO.20566 and the specific analysis in Reference 1, core cooling systems provide sufficient cooling to
~he t
core to dissipate the energy associated with theless-of coolant accident,tolimit calculated fuelcladdingte perature to less than 2200*F, to assure that core geometry remains intact, to limit cladding metal.waterie-~
action to less than 1%,and to limit the calculated local metal. water reaction to less than 17%.
.{ ,
The limiting conditions of operation in Specifications 3.5.A.1 through 3.5.A.6 specify the combinations '
of operable subsystems to assure the availability of the minimum cooling systems noted above. -
failure of ECCS equipment occurring during a loss-of-coolant accident under these limiting conditions of operadon will result in inadequate cooling of the reactor core.
l Core spray distribution has been shown, in full-scale iests of systems similar in design to that of Quad-Cities 1 and 2, to exceed the minimum requirements b) at least 25%. In addition, cooling cFectiveness has been demonstrated at less than half the rated flow in simulated fuel ass
_ beater. rods to duplicate the decay heat characteristies_of irradiated fuel. The accident analysis is additional conservative in that no credit is taken for spray cooling of the reactor core before the intema! ,'
preuure has fallen to 90 psis. o De LPCI mode of the RilR system is designed to provide emergency cooling to the core by flooding in l the event of a loss of-coolant accident. This system functions in combination with the core spray system to prevent excessive fuel cladding temperature. The LPCI mode of the RHR2 system in combinatio the core spray subsystem provides adequate coolinj for break areas of approximately 0.2 ft up to and including 4.18 ft , the latter being the double ended recirculation line break with the equalizer line 2
~
between th]'recirctilation' loops closed without assistance from'the high p'ressure emergency core cooli i
subsystems. '
' De allowable repair times are established so that the average risk raie for repair would be no greater than the basic risk rate.The method and concept are described in Reference 3 }Jsing the results developed inl this reference, the repair period is found to be less than half the test interval. This assumes that the <. ore spray subsystems and LPCI constitute a one out of-two system: however, the combined efTect of th systems to limit exceuive cladding temperature must also be considered. The test interval specifie; ..
- Specification 4.5 was 3 months. Therefore, an allowable repair period which maintains the' basic ris considering single failures should he less than 30 days, and this specification is within this period. For i
multiple failures, a shorter intervalis specified; to improve the assusance that the remaining systems(
function, a daily test is called for. Although it is recognized that the information given in Reference I l provides a quantitative method to estimate allowable repair times, the lack of opera ng data to supp t
' the analytical approach prevents complete acceptanec of this method at this time.Derefore, the times I stated in the specific items were established with due regard to judgment.
Should one core spray subsystem become inoperable, the rema'ning core spray subsystem and the en .
LPCI mode of the RHR system are available should the need fu core cooling arise: To assure that the
> . remaining core spray,:he LPCI mode of the RilR system, and the diesel generators are available, they '
are demonstrated to be operabic immediately.This demonstration includes a manual initiation of the l
pumps and anociated valves and diesel; enerators. Based on judgments of the reliability of th f systenis,i e., the core spray and LPCI, a 7. day repair period was obtained. -o l ! ,
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DPR-29 l H. Condensate Pump Room Flood Protection Sec Specification 3.5.H.
I. Aterage Planar I.HGR This specification assures that the peak cladding temperature following the postulated design-basis lossof<oolant accsdent will not exceed the 2200*F limit specified in the 10 CFR 50 Appendix K consdenng the postulated effects of fuel pellet densification.
The peak cladding temperature following a postulated lossof<oolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is only secondarily dependent on the rod to rod power distribution within an assembly. Stnce expected local variations in power distribution within a fuel assembly affect 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 LHCR is suf.
ficient to assure that calculated temperatures are below the limit. The maximum average planar LHCR's shown in Figure 3.51 are based on calculations employing the models described in Reference 2.
J. local I.HGH f This specification assures that the maximum linear heat generation rate in any rod is less than the design i linear heat. generation rate even if fuel pellet densification is postulated. The power spike penalty and assumes a linearly increasing variation in axial l Js disciassed in Reference 2 gaps between core bottom and top and assures with a 95L confidence that no more than one fuel red enceeds the design linear heat generatien rate due to power spiking.
l K. Minimum Critical Power Ratio (MCPR) i The steady state values for MCPR specified in this specification were selected to provide margin to accommo.
dat'e transients and uncertainties in monitoring the core operating state as well as uncertainties in the critical j power correlation itself. These values also assure that, operation will be sach that the initial condition assumed for the LOCA analysis plus two percent for-uncertainty is satisfied. For [
say of the special set of transients or disturbances caused by single ogrator error or snele equipment malfun: tion,it is required that deren analyses inmaliact at this steady state operating hmit ywid a MCPR of not ku thsn that s:ccified in Specifi:sti:n 1.1.A at so tirne during the tranuerit assuming mstrument tnp cett: ;s gMn m Specification 2.1. For ans.lysa of the thercal consequences of these transients, the value of MCPR stated in this 1 specification for the limiting condition of operation bounds the initial value of MOPR assumed to exist prior to the initiation of the transients.
This initial condition, which is used in the transient analyser, will cre-clude violation of the fuel cladding inte6rity safety limit. Ass unp tion.:
and methods used in calculating the recuired steady state MOF3 li-:it fer each reload cycle are docu .ented in Reference 2. The results apply with increased conservatism while operating with MCFR's greater than specified.
. The most hmiting transients with respect to MCP 7, are generaDy: .
a) Rod withdrawa!crror .
- b) Lead rejection or Turbine Trip without bypasa c) Im s of feedwater heates ,
Severa; factors influence which of these transients results in the lat. est reduction in critical power ratio such as the specific fuel loading, ex-posure, and fuel type. The current cycles reload licensing analyses spec-1 ifies the litniting transients for a given exposure increment for each fuel type. Tr.c values specified as the Liraiting, Condition of Operation are con-servatively chosen to t)ound the most restrictive over the ent. ire cyef e for l each fuel type. . t i I 3.V4 .%-14 I i.
- Qt!A D. CITIES
.DPR- 29 For cose flow sates less than rated. t!ie steady itste hfCPR is increased by the forrnula given in the i,'ce#,
sation. h suures that the htCPP wdl be mamismed treates than thst speci6cd in Specification I I A eicn in the evesit sfiat t!.e motor teneratos set speed contro!!ct esuses the scoop tube poutioner for the lhJ c.,;;Ier to move to the maaimum speed positson.
References .
- 1. " Loss-of-Coolant Analysis' Report for Dre, den Units 2, 3 and Quad Cities Units 1, 2 Nuclear Power Stations," HEDO-211146A',
April, 1979
- 2. " Generic Reload Fuel Application," NEDE-240ll-P-A'*
- 3. I." M. Jacobs and Guidelines P. ~4. Marriott, for Determinine GE Topical Report APED S736, Safe Test Int.crvals and Hepair Ta.u for Er.gineered Safeguards," April,1969
- Approved revision at time of plant operation.
- Approved revision number at time reload fuel analyses ara ,
pe rfor.T.e d .
i .
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f 1
5 l
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, QLIAD CITIF.S DPR-29 l- ,
f3 V Should the switches at levels (a) and t h) f.sil or the operator fail to trip the circunting water pumps on alarm at level (b), the actuation of cather level switch pair at level (c) shall trip the circulatine water pumps automatically i
and alarm in the control room These redundant lesel swath pairs at level (c) are designed and installed to IH.1:
279,' Criteria for Nuclear Power Plant Protection Spreme As the circulating water pumps are tripped. cubes manually or autom.ttically, at level (c) of 5 feet. the masimum water level reached in the condenscr pit due to pumping will be at elevation $65 feet 6 inches elevation (10 feet above condenser pit floor elevation 558 feet 6 inches; 5 feet plus an additional 5 feet attributed to pump coastdown).
In order to present the RiiR service water pump motors and diesel-generator rooting water pump motors from overheating. a vault cooler i? supplicd for cach pump hact vault cooler is desirned to maintain the vault at a maximum 105
- F temperature durmg operation or m respective pump. I or esample. if ditsel gene ator conhnr water pump 1/2 3903 starts,its cooler also starts and mamtains the vault at 105' F by removmg heat supphed to the vault by the motor of pump I/2 3903. If,at the same time that pump 1/2 3903 is in operation. RilR seivite water pump IC starts. its cooler will also start and compensate for the added heat supplied to the vault 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 l cto 5% of the design flow of the purnps 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 Verification that acceo doors to each vault are closed following entrance by personnel is covered by station operating pruceduret The LHGR shall be checked daily to determine if fuel burnup or control rod movement has caused changes in power distribution. Since chances due to burnup are slow and only a few control rods are moved daily, a daily check of power distribution is adequate.
Aserage Planar LilGR At core thermal power levels less than or equal to 25%. operating plant experience and thermal hydraulic analyses mdicate that the resuhing average planar LilGR is below the maximum average planar 1IIGR by a considerabic margin; therefore, evaulanon of the average planas I.IIGR below this power level is not necewary. The daily requirement for calculating aserare planae LilGR above 25% rated thermal powet is suiheient, smcc power distribution shifts are slow when there have not been signiticant power or control rod changes.
Local LilGR The LHGR as a function of core height sh'all he checked daily during reactor operation at greater than or equal A
to 25% power to determine if fuel burnup or control rod movement has caused changes in power distribution.
limiting LHGR value is precluded by a considerable margin when employing any permissible control rod pattern below 25% rated thermal power.
Minimum Critical Pnact Itatio (MCPR)
At core thermal power levels less than or equal to 25%, the reactor will be operating at minimum recirculation pump speed and the moderator void content will he very small. For all designaicd control rod patterns which may be employed at this point, nperatinp plant emperience and thermal hydraulie analym indicate that the resulting MCPR value is in excess of requirements by a considerable margm. With this low void content, any inadvertent core flow inucase would only plate ope ation in a more conservative mode relative to MCPR.
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