ML20039G292
| ML20039G292 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 12/23/1981 |
| From: | Ippolito T Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20039G293 | List: |
| References | |
| NUDOCS 8201180008 | |
| Download: ML20039G292 (21) | |
Text
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'g5 UNITED STATES y3 NUCLEAR REGULATORY COMMisslON s
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... j WASHINGTON. D. C. 20555
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0, COMMONWEALTH EDISON COMPANY s'Tf AND ICWA-ILLINOIS GAS AND ELECTRIC COMPANY
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1 DOCKET NC. 50-265 Tsv:
00AD CITIES STATION UNIT N0. 2 f):i AMENCMENT TO FACILITY OPERATING LICENCE Amendcent N'o. 69 License No. DPR-30 1.
The Nuclear Regulatory Commission (the Commission) has found that:
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A.
The application for amendment by the Commonwealth Edison Ccmpany (the licensee) dated July 27, 1981, as supplemented August 21, 1981 and December 3,1981, complies with the standards and requirements of rules and regulations set forth in 10 CFR Chapter I;the Atomic E B.
The facility will operate in confomity with the application, the provistens of the Act, and.the rules and regulations of the Comisison; C.
There is reasonable asserance (1) that the activities authorized by this atencrent can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliarce with the Ccmmission's regulations; 3.
The issuance of this atencment will not be inimical to the common defense and securi:y or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amend-ment, and paragraph 3.B of Facility Operating License No. DPR-30 is hereby amended to read as follows:
A B.
Technical Specifications
>h The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 69, are hereby incorporated in the license.
The licensee shall oprate the facility in accordance with the Technical Specificati.ons.
8201180008 811223 PDR ADOCH C5v00265 P
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s 3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATOR' COMMISSION Thomas A. Ippolito, Chief Operating Reactors Branch #2 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: December 23, 1981 k
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4 ATTACHMENT TO LICENSE AMENDMENT NO. 69 FACILITY OPERATING LICENSE NO. DPR-30 DOCKET N0. 50-265 Replace the following pages of the Appendix "A" Technical Specifications with'the enclosed pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.
1.1/2.1-4 1.1/2.1-5 1.1/2.1-6 1.1/2.1-7 1.1/2.1-7a (new page) 1.1/2.1-11 1.2/2.2-1 1.2/2.2-2 1.2/2.2-2a (new page) 3.3/4.3-5 3.3./4.3-10 3.5/4.5-10 3.5/4.5-14 3.5/4.5-14a 3.5/4.5-14b Figure 3.5-1 (sheet 1 of 5)
Figure 3.5-1 (sheet 2 of 5)
Figure 3.5-1 (sheet 4 of 5)
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N QUAD-CITIES DPR-30 1.1 SAFETY LIMI? 19S13 The f' vel cladding integrity linit is set such that no calculated fuel oveeje sould necur as a recu1* er en abnormal operat aonal tr nsas nt, recause f uel damage is not directly ebaerveble, a stup back apps ocel. is used to establish a safety linit such that the minimum critical power ratio (ncen) in no less th.in the fuel cladding integrity safety linit Ecru > the fuel claddis,g integrity safety linit representa a conr.ervative margin relative to the conditaons required to maintain f usi cladding integrity.
The fuel cladding is cne of the physical boundaries whicts separate radioactive satertabs fran tne ertvirons.
The integrity of the fuel claddin6 is related to ita relative freedan frum perforations Or Cracking.
Although some corrosion or ume-related cracittng may occur during the life of the cladding, fissien pseduct migration from this source is inerencntally cut-Jintive and continuously measurabic. Fuel claddang m.
forations, however, can result from ther-al stscrees shich occur fro;n scactor operatien significantly al,nve design conditions and the protection systen safety settings. Whilu f assson eroduct migration tres eldden; perforation is jutt &s measurable as that frem u*c-related crackirq. the therually c uted cladds:4 uer<os-ations sienel a thresholc; beyond shach still greater ther-r.a1 stresses may cause ga ost ratPer than ancrement-al cladding deterioration. Therefore the fuel eladdang safety limit is duf tned with smarga n to the cen s-a tions shich would produce onset of transit aan boiltng (nCFR of 1.0). These conditions, represent a signift.
eent departure frcra the condition intended by design for planned operation. Therefore, the fuel claddtag integrity safety limit is established such that no calculated fuel denage shall result from an abnormal operational trancaesit. Sasas of the values derived for this safety limit for each fuel type is documented in Referencesi e.nd 2.
l A.
Raaetor Pressure > 800 psig and Core riow > 1C% of Rated enset of transition boiling results in a decrease in heat tran=fer from the claddiant and therefore elevated claddang teeperrture and the posnaoality of claddir.g failusc. IIowever, the existence of critical power, or boalang transitaon la not a directly observablu (v.remeter in an operetarwe react-or, therefore, the margan to boala g tranc.ition is calculated frem plant opur.:ta ng parrmeters such as core power, core (1cn*, feeGeter tenperature, and core power distr abut ta.n.
The margin f or each h-fuel assembly is characterized by the critical pcuer ratio (cru), shash as the ratio of the bar.dle pcuer which would produce onnet of transition boiling divided by the actual Lundle power. The miniacs value of this ratio for eny In hdle in the core is the rninimum ca itical pswer ratio (:4 rn).
It is assumed that the plar.t operatien is coetrolled to the ncn*inal protectave tresponts via the lastrumented variables (rigure 2.1-3).
The McPR fuel claddirry integrity sa'ety 11ait has suf ficient conaervatismi to asi,vre that in the event of an ebnormal operatsonal transient initiated frein the norruel oper.itarve conditivu. =cre taan 29.?z of the fuel rods in the core are expoeted to avond boiling transit ion.
The margi?i betscen rcPa of 1.0 (or. set of transation boilf rig) and the safety limit, is derived f rc:a a d: tailed statistacal analysis considering all of the uncertcinues in sionitoring the ccre operat snq state. ir.cluding uncertainty in tFe boiling transition correlation (see e.g.. keference 1).
Because the bostar.g transition correlation is based on a latte quantity of full-scale data, the re is a vray hagh con.
fidence that operatton of a fuel assembly at the condition of HCP*t = the fuel cladding integrity safety limit would not produce boiling transition.
Mowever, if boiling transition were to occur cladding perforation would not be expectawi. C3 c od t rw) temperatures would inerence to appronisntely 11CC*r. which is below the perferation t:1npernstr e of the cladding material. This han been verafied by tests in the cenarn) Electric test Reactor (OrtA).
where sirsilar fuel operated above the cratacal he.cn flux for a significant period of tamc (3C asan-utes) without claddig parforataan.
If reactor pressure should ever exceed 14C0 psia during nor r.at power operation (the limit of applicability of the boiling transition correlatica), it would be assured that the fuel cladding integrity safety limit has been violated.
2n addition to the boiling transition limit 4 cPR) operation is constrained to a maximum LH CRs17.5 kw/f t for 7 x 7 fuel and 13.4kw/f t for all 8x8 fuel types. This constraint is estabitshed by strain for abnorma. operating transients ini: yin to, rom nighNAStic mar lis specification 3.5.J.
to yrovide adeCuate SafetV
_atec.
power conditions.
Specification 2.1.A.1 provides for equivalent safety margin for transients initiated from lower power con-ditions by adjusting the APMi flow-biased scram setting by the ratio of FRP/MFLPD.
C Amendment No. F, 69 1.1/2.1-4 h
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Q(IAD-CllllS DPR-30 i
Specification 3.5J established the LilGR masimum which cannot he cacceded under steady power operation.
B.
Core Thermal Power Limit (Reactor Pressure <800 psia)
At pressures below 500 psia. stie core elevation pressure drop (0 pomer.0 flow)is greater than 4.56 psi.
At low powers and flows this pressure differentut is maintained in the bypats region of the cere. Since the pressure drop in the bypsss region is essentially all elevation head, the core pressure drep at low powers and flows will always be greater than 4.56 psi. Analyses show that with a flow of 23 x 10':b/hr bundle flow, bund:e pressure drop is nearly independent of bund!e power and has a value of 3.5 psi.'thus the bundle flow with a 4.56-psi driving head will be greater than 2R x 10'Ib/hr. Fu!! scale ATLAS test data taken at pressures from 14.7 psia to 600 psia indicate that the fuel assembly critics! power at this flow is approximately 3.35 MW. At 25% of rated thermal power, the peak powered bundle would have to be operating at 3.86 times the average powered bundle in order to achieve this bundle power. Thu>.
a core thermal power limit of 25% for reactor pressures below 800 psia is conservative.
C.
Power Transient During transient operation the heat flux (thermal power.to-water) wculd lag behind the neutren Sus due to the inherent heat transler time con > tant of the ruel. which is 8 to 9 secomis. Also, the limiting safety system scram settings are at values which wit! not allow the reactor to be operated above the safety limit (a
during r.ormal operation or during other pla..t operating situations w hath luve been analyzed in deta:1.
In addition, control rod scrams are such that for normal operating transients. the neutron Aux transient is terrninated before a significant increase in surface heat flux occurs. Control rod scram times are checked as required by Specification 4.3.C.
and the MCPR operating limit is modified as necessary per Specification 3 5.K.
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Exceeding a neutron flux scram setting and a failure of the control rods to reduce flux to less than the scram setting within I.5 seconds does not necessanly imply tnat fuel is damaged; however. for this specification, a safety limit violation will be assumed any time a neutron flus scram setting is exceeded for longer than 1.5 seconds.
If the scram occurs such that the neutron flux dwell time above the limiting ssfety system setting is less than 1.7 seconds, the safety limit will not be exceeded for normal turbine or generator trips, which are the mest severe normal operating transients expected.These ana!yses show that even if the bypasuistem fails to operate, the design limit of MCPR - the fuel cladding intecrity safety
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limit is not exceeded.
Thus, use of a 1.5 second limit provides additional mh. rein.ef has a sequence annunciation prograr6 m hich willindicate the sequence m. which The computer prov scrams occur, such as neutren f!ux. pressure, etc This pregram also indicates when the scram setpemt is cicated. This will previde information on how long a scram condition exists and thus provide some l
measure of the energy added during a transient.Thus, computer information normally will be available for snalyzing scrams; however, if the computer information should not he available for any scram analysis. Specification 1.1 C.2 will be relied on to determine if a safety limit has been violated.
i During periods when the reactor is shut down, consideration must also be given to w::ter level requirements due to the effect of decay heat. lf reactor water Icvel should drop below the top of the active fuel during this time, the abihty to cool the core is reduced. This reduction in core-coolin;; capability cou!d lead to elevated claddmp. temperatures and etaddmp rerferation.The enre will he evoted wii+ciently to prevent c!sdding mehing should the water level be reduced,to two-thods the noe height 1.st.ibbsh.
ment of the safety limit at 12 inches above the top of the fuel provides adequate marpin.'lhis level will be contionously monitored whenever the rmtrulanon rumps are not ogwratmp..
- Top of the active fuel is defined to be 360 inches above vessel zero (see Bases 3.2).
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Amepdment No. J(, 69
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Qt' AD-CIT 1i'S DPI(. \\0 References 1.
" Generic Reload fuel Applications," NEDE-24011-P-A*
n,uener e gni sr:r.a t, ion Fu r Ba rrier Fuel De:nens tra tion Eund '*e l
c.
,icensing', NEDO-24259-A, February 1981.
L
- Approved revision number at time reload fu:i analys'es are p'erforced.
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- 1.1/ 2.1 - 6
/ceneent no. pr, 69 l
Quad Cities DPR-30
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2.1 LIMITING SAFETY SYSTEM SETTING BASES Ine abnormal operational transients applicable to operation of the units nave been analyzed throughout the spectrum of planned operating conditions up to the rated thermal power condition of 2511 MWt.
In addition, 2511 MWt is the licensed maximum steady-state power level of the units.
This maximum steady-state power level will never knowingly be exceeded, g
Conservatism incorporated into the transient analysis is documented in References 1 ano 2.
Transient analyses are initiated at the ccnditions given in these References.
The scram delsy time and rate of rod insertion allowed by the analysesLare conservatively set equal to the longest delay and I
slowestrinsertion rate acceptable by technical specifications.
The effects.of scram worth, scram delay time, and rod insertion rate, all conservatively applied, are of greatest significance in the early portion of the negative reactivity insertion.
The rapid insertion of negative reactivity is assured by the time requirements for 5% and 20% insertion.
By the time the rods are 60% inserted, approximately 4 dollars of negative reactivity have been inserted, which strongly turns 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 establish the ultimate fully shutdown ste&dy-state condition.
The MCPR operating limit is, however, adjusted to account for the statistical variation of measured scram times as discussed in Reference 2 and the bases of Specification 3.5.K.
Steady-state operation without forced recirculation will not be permitted except during star tup testing.
The analysis to support operation at various power and flow relationships has considered operation with either one or two recirculation pumps.
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The bases for individual trip settings are discussed in the following paragraphs.
r For analyses of the thermal consequences of the transients, the MCPR's stated in Paragraph 3.5.K as the limiting condition of operation bound those which are conservatively assumed to exist prior to initiation of the transients.
A.
Neutron Flux 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 ratsd thermal power.
Because fission chambers provide the basis input signals, the APRM system responds directly to average neutron flux.
During transients, the instantaneous. rate of heat transfer from the fue'l (reactor thermal powerP*is less than the instantaneous
-neutron f. lux dua to the time constant of the fuel.
9mendment Nc. JT,'69
~1 1/2 1-7
Therefore, during abnormal operational transients, the thermal power of the fuel will be less than that indicated by the neutron flux it the scram setting.
Analyses demonstrate that with a 120% scram trip setting, none of the abnormal operational transients analyzeo violates the fuel safety limit, and there is a substantial margin from fuel damage.
Therefore, the use of flow-referenced scram trip provides even additional margin.
Amendment No. 69 1.1 2.1-7a a
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QUAD-CITIES DPR-30 References 1.
" Generic Reload Fuel Application," !!EDE-24011-P-A*
- Approved revis' ion number at the reload analyses are performed
- 2. " Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors" General Electric Co. Licensing Topical Report NEDO 24154 Vols. I and II and NEDE-2Lil54 Volume III ac supplemented by letter dated September 5,1980 from R. ii.
Euchholz (GE) to P. S. Check (NRC).
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Amendment No ff, 69 I 'Y *I""
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QtrAI)-CITIFS i
DPR-30 1.2/2.1 REACTOR COOLANT SYSTEM LIMIT 1NG.9AFETY SYSTF.M SF.TTING SAFETY LIMIT Applicability:
Appliesbility:
Applies to trip settings of the instruments and Applies to limits on re:eter coolant system devices which are provided to present the reactor pressure.
system safety limits from being eseceded.
Objective:
Objectlie:
To establish a limit below which the integrity of the To def.ne the level of the process variables at which reactor coolant system is not thie.ttened due to an automatic pro:ective action is initiated to prevent the safety limits frem being exceeded.
overpressure condition.
SPECIFICATIONS A.
Reactor coolant high. pressure scram sha!! be A.
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Primary system safety valve nominal settings shall be as fo'!cws:
I valve at 1115psis'"
2 valves at 1240 psig 2 valv:s at 1250 psig 4 valves at 1260 psig tirTarget Rock combinatinn safety / relief vahe The allowable setpoint error for each valve shall be i15 t
/cendmentNo.JT,69 yam e--
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QUA0 CITIES DPR-30 1.2 SAFETY LIMIT BASES The reactor coolant systen integrity is an important barrier in the prevention of uncontrolled release of fission products.
It is essential that the integrity cf this system be protectsd by establishing a pressure limit to be observed for all operating conaitions anc whenever there is irradiated fuel in the reactor vessel.
The pressure safety limit 1345 psig as measured by the vessel l
steam space pressure indicator is equivalent to 1375 psig at the lowest elevation of the reactor vessel.
The 1375 psig value is l
derived from the design pressures of the reactor pressure vessel and coolant system piping.
The respective design pressures are i
1250 psig at 5750F and 1175 psig at 5600F.
The pressure i
safety limit was chosen as the lower of the pressure transients permitted by the applicable design codes.
ASME Boiler and Pressure Vessel Code Section III for the pressure vessel, and USASI 831.1 Code for the reactor coolant system piping.
The ASME Boiler and Pressure vessel Code permits pressure transients up to 10% over design pressure (110% x 1250 = 1375 psig), and the USASI Code permits pressure transients up to 20% over design pressure (120% x 1175 = 1410 psig).
The safety limit pressure of 1375 psig is referenced to the lowest elevation of the I
reactor vessel.
The design pressure for the recirc. suction line piping (1175 psig) was chosen relative to the reactor vessel design pressure.
Demonstrating compliance of~ peak vessel pressure with the ASME overpressure protection limit (1375 psig) assures compliance of the suction piping with the USASI limit (1410 psig).
Evaluation methodology to assure that this safety limit pressure is not exceeded for any reload is documentea in Reference 1.
The design basis for the reactor pressure vessel makes evident the substantial margin of protection against failure at the safety pressure limit of 1375 psig.
The vessel has been designea for a general membrane stress no greater than 26,700 psi at an internal pressure of 1250 psig; this is a factor of 1.5 below the yiela strength of 40,l00 psi at 5750F.
At the pressure limit of 1375 psig, the general membrane stress will only be 29,400 psi, still safely below the yielo strength.
The relationships of stress levels.to yield strength are comparable for the primary system piping and provide similar margin of protection at the established safety pressure limit.
The normal operating pressure of the reactor coolant system is 1000 psig.
For the turbine trip or loss of electrical load transients, the turbine trip scram or g'enerator load rejection scram together with the turbine bypass system limits pressure to approximately 1100 psig (References 2,3, and 4).
In addition, pressure relief valves have bean provided to reduce the probability of the safety'valver operating in the event that the turbine bypass shoulo fail.
1,.2/2.2-2 7:
Amendment No.[, 69
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Finally, the safety valves are sized to keep the reactor vessel peak pressure below 1375 psig with no' credit taken for relief valves during the postulated full closure of all MSIVs althout direct (valve position switch) scram.
Credit is taken for the neutron flux scram, however.
The indirect flux scram and safety valve actuation provide adequate margin below the allowable peak vessel pressure of 1375 psig.
4 Reactor pressure is continuously monitored in the control room during operation on a 1500 psi full-scale pressure recorder.
References
- 1. " Generic Reload Fuel Application," NEDE-240ll-P-A*
- 2. SAR, Section 11.22
- 3. Quad Cities 1 Nuclear Power Station first reload license i
submittal, Section 6.2.4.2, February 1974
- 4. GE Topical Report NE00-20693, General Electric Boiling Water Reactor No. 1 Licensing submittal for Quad Cities Nuclear Power Station Unit 2, December 1974.
i Approved revision number at time reload analyses are performed.
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1.2/2.2-2a Amendment No. 69 i
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n QUAD-Cli!ES DPR-30 sicered inoperable, fully provide reasonable assurance inserted into the core, that preger control rod drive and electrically disarmec, performance is being maintatnec. The results of measurements performed on the 5.
If the overall average control roo crives snali ce of tne 20% insertion scram submittec in the annual operating time cata generatea to report to the NRC.
Cate in the current cycle exceeds 0.73 seconos, the MCPR operating limit must 5. The cycle cumulative mean be modified as required by scram time for 20s insertion Specification 3.5.K.
will be determined immediately following the testing requirec in Specifications 4.3.C.1 and 4.3.C.2 and the MCPR operating limit adjusted, if necessary, as reovired by Specification 3.5.K.
D. Control Rod Accumulesors D. Comsrol Rod Accumulators At all reactor operatmg pressures, a rod accu.
Once a shift. check the status of the pressure mulator may be inoperable provided that no and level alarms for each atrumulator other control rod in the nine rod square array around this rod has' l.
- 2. a directional control valve electncally disarmed while in a nonfully mserted posinon. or
- 3. a scram insertion greater than man-imum permissable insertaon time.
If a control rod with an inoperable accumulator
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ts insened full-m and its darectional control valves are electrically disarmed. at shall not be considered to have an anoper.bie accumulator.
and the rod block associated with that inopera-ble accumulator may be bypn ed E.
Renetivity Anemehes E.
Remetivity Amomehes The reactivity equivalent of the diference Durmg the startup test program and startups between the actual cntical rod con 6gurauon following refuehng ouiages. the crsucal rod and the espected configurataan dunng power configuranons =ill be compared to the expected operauon shall not exceed IMk. If this hmit is configurauons at selected operaung condtuons.
exceeded, the reactor shall be shutdown until These cornpansons will be used as base data for the cause has been determined and corrective reactmty monitormg durmg subwquent power actions have been taken. In accordance with operauon throughout the fuel cycle. At spec:6c Spec 26 cation 6.6, the NRC shall be notined of power operaung conditions. the crtucal rod tha reportable occurrence wuhs: 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
con 6guration will be compared to the conhg-urauen expected based upon appropnately cor-rected past data.This companson will he made at least every equivalent full power month.
F.
Economic Generstion Control Setem F.
Econnaie Generosion Coneral splem Operation of the unit with the economic gener-The range set into the economic generation anon control system with automauc f!ow con-control system shall be recorded weekly trol shall be permassable only in the range of 65% to 100". of rated core flow, with reactor power above 20%
3.3/ O-S Amendment No. 69
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QUAD CITIES
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o C.
Scram Insertion Times The control rod system is analyzed to bring the reactor suberitic tl at I
a rate fast encuch to prevent fuel damage, i.e., to orevent the MCPR frc: becoming less than the fuel cladding integrity safety limit.
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nnalysis lif'the' limiting power.
transient sacks that the negative reactivity rates resulting frc= the scram with the average r:27 Me-all the drives as given in the above specification, provide the required protection, and MCFR re=ains greater than the fuel cladding inteirity safety limit. It is necessary to raise the MCPR operating limit (per Specification 3 5.K) when the average 20% scram insertion time reaches 0.73 seconds on a cycle cumulative basis (overall average of surveillance data to date) in order to comply with assumptions in the implementation procedure for the ODYN transient analysis computer code.
The basis for choosing 0.73 seconds is discussed furtner in the bases for Specification 3 5.K.
In the analytical treatment of the transients, 290 milliseconds are allowed between a neutron sensor reaching the scram point and the start of motion of the control rods.
This is adequate and conservative when compared to the typically observed time delay of about 210 milliseconds.
Approximately 90 milliseconds af ter neutron flux reaches the trip point, the pilot scram valve salenoid deenergizes and 120 milliseconds later the control rod motion is estimated to actually begin.
However, 200 milliseconds rather than 120 milliseconds is conservatively assumed for this time interval in the transient analyses and is also included in the allow-able scram insertion times specified in Specification 3 3.C.
The scram times for a!! control rods wi!! be determined at the time of each refueling outage. A representative sample of control rods will be scrim tested following a shutdown.
Scram times of new drives are approximately 2.5 to 3 seconds; lower rates of change in scram tirnes fo!!owing initial plant operation at power are expected.
The test schedule prevides re.tsanable assurance of detection of slow drives before system deterioration beyond the limits of Specifintm 3.3.C. The pregram was deve!eped en the basis of the sta:Istical approach out!!ned belcw
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and judgment.
The history ordrive performance accumulated to date indicates that the 90% in<ertion times of new and overhau!ed drives approximate a normal distribution about the mean which tends to become skewed toward longer scram times as operating timeis accurnulated.The probability of a drive not exceeding the mean 90% insertion time by 0.75 seconds is greater than 0.999 for a normal distribution. The measurement of the scram performance of the drises surtnunding a drive exceeding the expected range of scram performance will detcet local variatier.s and also provide assurance that local scram time !imits are not exceeded. Continued monitoring of other drives exceeding the expected range of scram times provides surveillance of possible anomalous performance.
The numerical values assigned to the predicted scram performance are based on the analysis of the Dresden 2 startup data and of data from other BWRi such as Nine Mile Point and Oyster Creek.
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The occurrence of serain times within the limits, but significantly longer than average, should be viewed as an indication of a systematic prehlem with control rod drives, especially if the number of drives exhibiting such scram times exceeds eight, the allowahic number ofinoperabic rods.
3.3M.3 - 10 Amendment No. Jf, 69
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DPR 70 CU.cD CITIM ITUT* 2
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Rate (MAPLEGR) m.
Planar Average Exposure
=
Amendment No. 34, 69
QUAD CITIES DPR-30 ~
within the presented Lmi:s within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. the reset rst shaU be bicught to th: cc;d shutdown cond tion within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Sarvedlance and cor-respondmg a: tion stan contmue untd resetor opera:2on is wnhia the preset: bed hmits.
Maximum allowable LHGR for all 8X0 ruel types is 13.h KW/ft.
f K.
Minimum Critical Power K.
Minimum Critical Power Ratio (MCPR)
Ratio (MCPR)
During s teady-sta te opera tion The MCPR shall be determined
[
at rated core flow, MCPR shall daily during steady-s ta te -
be greater than or equal to:
power operation above 25%
of rated thermal power.
1 37 for T 5 0.73 secs 3ye 1.h2 for T ve 2 0.86 secs a
0 385 Tave + l 089 for o.73 < Tave < o.86 secs where $ve = mean 20fo sc ram insertion time for all surveillance data from Scecification h.3.C. whicEl has oeen generated in the current cycle.
For core flows other than rated, these nominal values of MCPR shall be increased oy a factor of ke
'4here kr is as shown in Figure 3 5.2.
f any time during operattori it is determined by normal surveillance that the limiting value for MCPR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits.
If the steady-sta te MCPR is not returned to 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 she cold shutdowri condition wfthin 30 Surveillance and correspond-hours.
ing action shall continue until reactor operatien is within the prescribed limits.
3.5/4.5-10 Amendment No. pf, 69
~
E e
QUA0 CITIES DPR-30 shown on Figure 3.5-1 as limits because conformance calculations have not been performed to justify operation at LHGR's in excess of those shown.
J.
Local LHGR This specification assures that the maximum linear heat-generation rate in any rod is less than the cesign linear heat-generation rate even if fuel pellet densification is postulated. The power sotke penalty is discussea in Reference 2 and assumes a 119early increasing nariation in amial ga;s between core bottom anc top anc assures with 95% confidence that no more than one f uel roc exceeds the design LHGR cue to power sotking. No penalty is required in Specification 3.5.L because it has been accounted for in the reload transient analyses by increasing the calculated peak iHGR by 2.2%.
K.
Minimum Critical Power Ratio (MCPR)
The steady state values for MCPR specifiec in this specification were selected to provice margin to accomodate transients and uncertainties in monitoring the core operating state as well as uncertainties in the critical power correlation it se lf. These values also assure that operation will be such that the intittal concition assumea for the LOCA analysis plus two percent for uncertainity is satisfied. For any of the special set of transients or disturDances caused by single operator error or single equipment malfunction, it is required that design analyses initialized at this steady-state operating limit yield a MCPR of not less than that specified in Specification 1.1.A at any time during the transtent, assuming instrument trip settings given in Specification 2.1.
For analysis of the thermal consequences of these transients, the value of MCPR stated in this specification for the limiting concition of operation bounds the initial value of MCPR assumec to exist prior to the initiation of the transients. This initial condition. which is used in the transient analyses, will preclude violation of the fuel cleading integrity safety limit. Assumptions and methods used in calculating the required steady state MCPR limit for each reload cycle are documented in References 2, 4, ana 5.
The l
results apply with increased conservatism while operating with MCPR's greater than specified.
The most limiting transtents with respect to MCPR are generally:
a) Roc withdrawal error b) Load rejection or turbine trip without bypass c) Loss of feedwater heater Several factors influence which of the these transients results in the largest reduction in critical power ratio such as the specific fuel loading exposure, and fuel type.
The current cycle's reloaa licensing analyses specifies the limiting transients for a given exposure increment for each fuel type.
The values specifica as the Limiting Condition of Operation are conservatively chosen to bound the most restrictive over the entire cycle for each fuel type.
The need to adjust the MCPR operating limit as a function of scram time arises from the statistical approach used in the implementatiJn of the ODfM computer code for analyzing rapid pressurization events. Generic statistical analyses were performed for plant groupings of similar design which considered the statistical variation in several parameters (initial power level. CRD scram insertion time, anc medel uncertainty). These analyses (which are describec further in Reference 4) produced generic Statistical Adjustment Factors which have been applied to plant and cycle specific 00fN results to yield operating limits which provide a 95%
probability with 95% confidence that the limiting pressurization event will not cause MCPR to f all below the fuel claccing integrity safety limit.
3.5/4.5-14 AmendmentNo.jFf, 69
=
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QUAD-CITIES DPR-30 As a result of this 95/95 approach, the average 20% insertion scram time must be monitoreo to assure compliance with the assumed statistical distribution.
If the mean value on a cycle cumulative (running average) basis were to exceed a 5% significance level comparea to the distribution assumeo in the 00YN statistical analyses, the MCPR limit must be increased linearly (as a function of the mean 20% scram time) to a more conservative value which reflects an NRC determined uncertainty penalty of 4.4%.
This penalty is soplied to the plant specific ODYN results (i.e. without statistical adjustment) for the limiting single failure pressurization event occurring at the limiting point in the cycle..
It is not applied in full until the mean of all current cycle 20%
scram tines reaches the 0.90 secs value of Specification
.3.3.C.I.
In practice, however, the requirements of 3.3.C.1 would most likely be reached (i.e. individual data set average >.90 secs) and the requireo actions taken (3.3.C.2) well before the running average exceeds 0.90 secs.
The 5% significance level is defined in Reference 4 as:
1/f.i Nj)l/2 e 7=4+ 1.65 (N 9
where Af = mean value for statistical scram time distribution to 20% inserted CF standard deviation of above distribution
=
N 1 = number of rods tested at B0C (all operable rods) nEN; = total number of operable rods tested in I'l the current cycle T h e v a l u e f o r 7"g u s ed in Specification 3.5.K is 0.73 secs which is conservative f or the f ollowing reasons:
a)
For simplicity in formulating and implementing the LCO, a conservative value for j$N i of 708 (i.e. 4x177) was used.
This represents one full core data set at BOC plus 6 half core data sets.
At the maximum frequency allowed by Specification 4.3.C.2 (16 week intervals) this is equivalent to 24 operating mor.ths.
That is, a cycle length was assumed which is longer than any past or contemplated refueling interval and the number of rods tested was maximized in order to simplify and conservatively reduce the criteria for the scram time at which MCPR penalization is necessary, b)
The values of #and CF were also chosen conservatively based on the dropout of the position 39 RPIS switch, since pos. 38.4 is the precise point at which 20% insertion is reached.
As a result Spec i f ica t ion 3.5.K initiates the linear MCPR penalty at a slightly lower v a l u e 7~a v e.
This alsu produces the full 4.4%
penalty at 0.86 secs which wou Td occur sooner th an the require'd value of 0.90 secs.
3.5/4.5-14a s
Amendment No. yT, 69
~
QUAD CITIES DPR-30 For core flow rates less than rated, the steady state MCPR is increased by the formula given in the specification.
This ensures that the MCPR will be maintained greater than that specified in Specification 1.1.A even in the event that the motor-generator set soeed controller causes the scoop tube positioner for the fluid coupler to move to the maximum speed position.
References 1.
" Loss-of-Coolant Analysis Report for Dresden Units 2, 3, and Quad Cities Units 1, 2 Nuclear Power Stations," NE00-2d146A*,
April, 1979 2.
" Generic Reload Fuel Application," NEDE-240ll-P-A**
3.
- 1. M. Jacobs and P. W. Marriott, GE Topical Report APED 5736,
" Guidelines for Determining Safe Test Intervals and-Repair Times for Engineereo Safeguards," April, 1969.
4
" Qualification of the One-0imensional Core Transient Model for Boiling Water Reactors" General Electric Co. Licensing Topical i
Report NEDO 24154 Vols. I and II and NEDE-24154 Vol. III as supplemented by letter dated September 5, 1980 from R. H.
Suchholz (GE) to P. S. Check (NRC).
5.
Letter, R. H. Buchholz (GE) to P. S. Check (NRC) dated January 19, 1981 "0DYN Adjustment Methods For Determination of Operating Limits".
i Approved revision at time of plant operation.
J l
Approved revision number at time reload fuel analyses are j
performed.
1 l
3.5/4.5-14b Amendment No. 69 4
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