ML20067E160

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Amend 83 to License DPR-29,incorporating Revised MAPLHGR Limits for Certain Fuel Types & Deleting Min Critical Power Ratio & MAPLHGR Operating Limits for All 7x7 Fuel
ML20067E160
Person / Time
Site: Quad Cities 
Issue date: 12/15/1982
From: Vassallo D
Office of Nuclear Reactor Regulation
To:
Commonwealth Edison Co, Iowa Illinois Gas & Electric Company
Shared Package
ML20067E163 List:
References
DPR-29-A-083 NUDOCS 8301030363
Download: ML20067E160 (23)
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{{#Wiki_filter:. j o aro g 4 70, UNITED STATES y g s e. (,' p, NUCl. EAR REGULATORY COMMISSION \\" E WASHINGTON. D. C. 20555 t 8 %, **....o', COMMONWEALTH EDISON COMPANf AND IOWA ILLIN0IS GAS T D ELECTRIC COMPANY DOCKET NO. 50-254 QUAD CITIES NUCLEAR POWER STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 83 License No. DPR-29 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment by the Commonwealth Edison Company (the licensee) dated August 19, 1982, as supplemented by two letters dated October 18, 1982, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the connon defense and security or to the health and safety of the public; and E. The issuance of this amencment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied. 2. Accordingly, the license is amended by changes to the Technical Spec-ifications as indicated in the attachment to this license amendment and paragraph 3.8 of Facility License No. DPR-29 is hereby amended to read as folicws: 8. Technical Scecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 83, are hereby incorporated in the license. The licensee shall operate the facility in accorcance with the Technical Specifications. 8301030363 821215 PDR ADOCK 05000254 P PDR

a 3. This license amendment is effective as of the date of issuance. FOR THE NUCLEAR REGULATORY COMMISSION Domenic B. Vassallo, Chief Operating Reactors Branch #2 Division of Licensing

Attachment:

Changes to the Technical Specificctions Date of Issuance: December 15, 1982 O 9 O

3 ATTACHMENT TO LICENSE AMENDMENT N0.83 FACILITY OPERATING LICENSE NO. DPR-29 00CKET NO. 50-254 Revise the Appendix "A" Technical Specifications as follows: Remove Replace 1.1/2.1-4 1.1/2.1-4 1.1/2.1-5 1.1/2.1-5 1.1/2.1-7 1.1/2.1 - 7 1.1/2.1-7a 1.1/2.1-11 1.1/2.1-11 1.2/2.2-1 1.2/2.2-1 1.2/2.2-2 1.2/2.2-2 1.2/2.2-2a 3.3/?.3-5 3.3/4.3-5 3.3/4.3-10 3.3/4.3-10 3.5/4.5-10 3.5/4.5-10 3.5/4.5-13a 3.5/4.5-14 3.5/4.5-14 3.5/4.5-14a 3.5/4.5-15 3.5/4.5-15 3.6/4.6-4 3.6/4.6-4 Fig. 3.5-1 Fig. 3.5-1 (6 pages) (4 pages) 1

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4. _m. ~. p k. / QUAD-CITIES DPR-29 l 1.1 SAFETY I.IMIT W SIS 4 The fuel cladding integrity linit is set such that no enleulated fvet d6=3ge would occue as a rese1* of en abnormal operational tr.nsient. liceause fuel damage to not directly ebaervable, a stup back afyseert. is osed to establish 'a sarety limit such that the mani=um critteel f ewer ratio (MCrn) is no Jess than the fuel eladding integrity safety linit.rrf'it > the fuel cladding integrity safety limit represcata a cosaservative margin relative to the conditions required to maintain fuel e1 adding integrity. "he fuel cladding ia one of the pnysical coundaries which separate radioactive asterials frse the environs. The integrity of tne fuel cladding ta related to its reistive freedas tras pectornations or cracking. Althogh some corrosion or ume-related cracktng oay occur during the life el the etadding, fissien product migration from this source is incrementally cur?J3ative and continuounty measurable. ruel Claddang per. forations, however, een result from thermal strerses which occur fress scactor operatien significantly above design conditions and the protection system safety settings. Ehile fiasso's product migration fres claddans perforation is just as measurable as that fres u*e related cracking, the thertaally c.uted claddisq geerfor. etions signal a threshold beyo:us whach still greater thermal stresses may cause gross retPer than incraxnt. al cladding deterioration. Therefore, the fuel eleddang safety lia.it is defined with asargin to the cends. tions shach would produce onset of transat son boalang (::CPR of 1.0). Thtse conditions represent a p a gniti. eent departure frcra the cor.dition intended by design for planned operation. Therefore, the fuel c1PfMthg integrity safety limit is estabitshed such that no calculated fuel danage shall result from an abnormal opcrataenal transaant. Sasas of the values derived for this safety laJait for each twel type is documented in Reference 1 A. Roseter preeeure > 8D0 psig and Core Flow > 10% of pated Onnet of transition Lolling results in a deeiresse in heat tramfer from the eleddirig and thernfore elevated claddang temperrture and the possacality of eladding failure. liowever, the entstence of ~ critical power, or boaling transation as not a directly observablu partnieter in an operatanef re.ict. or. Therefore, the margin to boilang trane ttien is calculateG freu plant oper.iting parrmeten s such as core power. core (1cm. feeGeter temperature, and core power distt abut een. The margin ts er et, fuoi assac6hly is ch:ractersted by the critical peer retto (Crk). khaca as tha ratio of the L4r. die power which would produce onset of transitaon boiling davaded by the actual laundle power. The nintimum value of this ratio for any bundle in the core is the e.inimum cratic.1 power ratio (:t*!'n), 14 is titumed that the plar.t oMration is coetrolled to the naminal protectave actponts via the instrumented variables (Figure 2.1 1). The itCPit fuel claddang integrity safety limit has suf ficient con.ervat'iam to asiure that 'in the even. of an abnormal operational transsent anistated free tne norm..) ope r n arwt c on;i t t lun. mere tasa 1'9. m of the fuel rods in the core are exps,cted to avond boiling transition. The margan betwea:n trPn of 1.0 (on=ct of ':ransitaon boilleg) and the safety limit. is derived f ro a e detailed statistacal analysim considering all of the uncertainties in aionitoring the core operetthq state, includang uncertainty in the besling transition costslation (see e.g., heference 1). Because the boaltog transitten correlation is based on a larTe eruantity of full-sente stata, there is a very hagh con. fidence that operatten of a fuel asacN:1y at the condition of itCim = the fuel cladding integr.ty safety limit would not produce boiling transition. However, if boiling transition were to occur, eledding perforation would not be espectre. Cleddi rw) temperatures would increate to appro-inntely 11CC'f. which is belcw the perforatnon tropurntuse of the cladding material. This hai bten verafied by tests in the cenern) r.lectrae Test Reactor (CTrt.), where sinitar fuel operated ebove the cratical heat flus for a significant perios of tame (1C man. utes) without cladding parforatton. If reactor pressure should ever exceed 1400 psia during normal power operation (the limit of applicability of thu boiling transition correlatico), it would be assumed that the fuel cledding integrity saf ety limit has been violated. In addition to the boiling transition limit (reps) operation is constrained to a maximum L M CRs17.5 kw/f t for 7 m 7 fuel and 11.4kw/f t for all ex8 fuel types. This constraant is estahl,1shed by specif tention 1.t.a. to arevide adequate safety margin to 1% plastic strain tor abnormal cperating transacnts in:: atec from n gh power conditions. Specification 2.1.A.1 provides for equivalent safety margin for transients initiated frem lower power con-ditions by adjusting the APM flow-biased scram setting by the ratio of TRP/MFLPD. C ~ ~ 1.1/2.i-4 l Amendment No, U, 83 l e

w r,_. ~ QtIA D-Cl lll'% npa-29 A Speci5 cation 3.5J established the LilGR maximum which cannot he exceeded under steady power operation. B. Core Thermal Power Limit (Reactor Pressurc<800 psia) At pressures below 500 psia, the core elevation pressure drop (C power 0 flow)is greater than 4.56 psi. At low powers and flows this pressure differentialis maintained in the bypass region of the cere. Since the pressure drop in the bypass region is essentially all elevation head. the core pressure drop at low powers and flows will always be greater than 4.56 psi. Analyses rhow that with a Dow of 28 x 10'Ib/hr bund!c flow, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi."Ihus the bundle f!ow with a 4 56. psi driving head will be greater than 2R x 10'Ib/hr. Full seale ATLAS test data talen at pressures from 14.7 psia to 800 psia indit;ste that the fuel assembly critical power at this flow is approximately 3.35 MWt. At 25% of rated thermal power. the peak powered bundle would have to be operating at 3.86 times the average powered bundle in o. der to achieve this bundle power. Thus, 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 behim! the neutron flux due to the inherent heat transfer time constant of the fuel, which is $ to 9 secomit Also. the limiting safety-system scram settingt are at values which will not allow the reactor to he operated above the safety limit ~% during normal operation or during other plant operating situations u hach hase been analyzed in deta:'. In addition. control rod scrarns are such that for normal operating transients. the neutron Aux transient is terminated befort a signi$ cant increase in surface heat flux occurs. Control rod scra:n times are checked as required by Specification 4.3.C. and the MCPR operating 11=1t is =odified as necessary per Specification 3 5.K. Exceeding a neutron flus scram seuing and a failure of the control rods to reduce flux to less than the scram setting within 1.5 seconds does not necessan!y imply that fuel as darnaced; howeyer. for ints specineation, a safety hmit violation will be assumed any ume a neutron Dux scram setting is exceeded for longer than 1.5 seernds. If the scram occurs such 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 ana!yses show that even if the byriass system l fails to operate, the design limit of MCPR = the fuel cladding intecrity safety I limit is not exceeded. Thus, use of a 1.5 second li.mit prov 5es add itional md !has a sequence annunesation pregrar6 which wn.!Iindicate the sequence in whith ~in. The computer pro T scrams occur, such as neutren f!ux, pre <sure, etc. This pregram also indicates u hen the scram setpeint is cleared. This will previde information on how long a scram eendition exists and thus provide some measure of the energy added during a transient. Thus, computer information normally will be availab!e for analyzing scrams; however, if the computer information should not he available for any scram analysis. Specification 1.1 C.2 will be relied on to Jetermine if a safety limit has been violated. During periods when the reactor is shut down, ennsideration must also be given to water level requirements due to the effect of decay heat. lf reactor water level should drop below the top of the active fuel during this time, the ability to cool the core is reduced. This tcJuetion in enre.enohng e.ipahihty coufd lead la elevated claddm3 temperatures and cladding perforation.The enre will he cooled wtsrienth to prevent sladding mehint should the water lesel be reduced,to two tima il e suae he'E t i st.shhsh. h l rnent of the safety limit at 12 inches atee the top of the fuel provides adequale m.ir;'m. 'lhis level mill I be continuously momimed whenever the terirrulanon ruinps are nni cperatmg

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

I.1/2.15 Amendment No. H, 83

..- - -. m.: u.......- - a -. = ..=a Quad Cities DPR-29 2.1 LIMITING SAFETY SYSTEM SETTING BASIS ~ The abnormal operational transients applicable to operation of the units have 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 conditions given in these References. g I The scram delay time and rate of rod insertion allowed by the analyses are conservatively set equal to the longest delay and I slowest insertion 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 collars of negative reactivity have been inserted, which strongly turns the transient and acco'mplishes 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 steady-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 startup testing. The analysis to support operation at various power and flow relationsnips has considered operation with either one or two recirculation pumps. The bases for individual trip settings are discussed in the following paragraphs. For analyses of the thermal consequences of the transients, the MCPR's stateo 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

1. APRM Flux Scram Trip Setting (Run Mode)

The average power range monitoring (APRM) system, which is calibrated using heat balance data taken during steady-state conditions, reads in percent of rated thermal power. Because 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 fuel (reactor thermal power) is less than the instantaneous neutron flux due to the time constant of the fuel. Amendment No. 67,. 83 1.1/2.1-7

... - -....-----.. --..:.. =...-....-.. 1. '.... : = ='X'. Therefore, during abnormal operational transients, the thermal power of the fuel will be less than that i ndicated by the neutron flux at the scram setting. Analyses demonstrate that with a 120% scram trip setting, none of the abnormal operational transients analyzed violates the fuel safety limit, and there is a substantial margin from fuel damage. Therefore, the use of flow-referenced scram trip provides even additional margin. l l \\ Amendment No. 83 1.1/2.1-7a m ,-,--,----,-s--, -c--- -,--e

..,.= ---.-. _ :- - 5 mk..ai:.xm .......w.._.-~. i QllAD.CITtES DPR-29 References 1. " Generic Reload Fuel Application," 2;EDE-24011-P-A* i

  • Approved revision number at time reload analyses are performed
2. "Qualificati:n of the One-Dimensional Core Transient Model for Boiling Water Reactors" General Electric Co. Licensing Topical Report IEDO 2h15h Vols. I and II and NEDE-2h154 volume III as supplemented oy letter dated September 5,19eC from R. H.

Euchholt (GE) ,0 P. S. Check (NRC). I ( 1.1/2.1-11 i Amendment No. 6J, 83

4, 3 ouAn-nTir.s DPR-29 4F 1.2/2.2 REACTOR COOLANT $YSTEM UMiT1NG SAFETY SYSTT.M SF.TTING SAFITY LIMIT Appticability: Applicability: Applies to trip settings of the instruments and Applies to limits on reietor coolant system devices which are provided in prevent the reactor preuvre. system safety limiu from being esteeded. Objwtive: 06Jeetive: To d Ane the level of the process variables at which 'To estabfish a limit bete-which the integrity of the automatic pretretive acnon is initiated to prevent resetor coo! ant system is not threatened due so an the safety limiu from being esceeded. arrerpreuure condinen. SPECIFICATIONS A. Resetor coc! ant high.preuvre seram sha!! be a.w..,.i.as.,.i..,,............,.., s.4'.;.u.':.:..l,O.':..J.:i.'='Jm;;;;; OFJ';,::;;;' s1060 psig. i B. Primary system safety vaht nominal settings shall be as follows: I valve at 1835 psig" l t 2 velves at 1240 psig 2 valves at 1250 psig 4 valves et 1260 psig I arTarget Rock combination safety / relief vaht The ettowable setpoint error for each valve shat!be i 15 8.2 d 1 Amendment No. 83 _. _ _. _ _ _ _ _ ~ _ _ _

.a _. ..-._c.- -.__c._ _. ;... m. ..;.._ - _ m ;_ u_: an as: ^ QUA0 CITIES D PR-29 s.. 1.2 SAFETY LIMIT BASES The reactor coolant system integrity is an important barrier in the prevention of uncontrolled release of fission products. It is essential that the integrity of this system be protected by establishing a pressure limit to be observed for all operating conditions and 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 1250 psig at 5750F and 1175 psig at 560 F. The pressure 0 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 USAS1 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), anc the USASI Cooe 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 reactor vessel. The design pressure for the recirc. suction line piping (1175 psig) was chosen relative to the reactor vessel design pressure. Demonstrating compliaace of peak vessek 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 documented in Reference 1. The design basis for the reactor pressure vessel makes evident the substantial margin of protection agai'nst l failure at the safety pressure limit of 1375 psig. The vessel has been designed 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 yield strenght of 40,100 psi at i 5750F. At the pressure limit of 1375 psig, the general membrane stress will only be 29,400 psi, still safely below the yiela 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 reactor coolant system is.1000 psig. For tne turbine trip or loss of electrical load transients, the turbine trip scram or generator load rejection scram together with the turbine bypass system limits pressure to approximately 1100 psig (References 2,3, and 4). In adcition, pressure relief valves have been provided to reduce the probability of the safety valves operating in the event that the turbine bypass should fail. 1.2/2.2-2 Acendment No gJ, 83

QUAD CITIES DPR-29 Finally, the safety valves are sized to keep the reactor vessel l peak pressure below 1375 psig with no credit taken for relief I valves during the postulated full closure of all MSIVs without 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. Reactor pressure is continuously monitored in the control room during operation on a 1500 psi full-scale pressure recorder. References

1. " Generic Reloac Fuel Application," NEDE-240ll-P-A*
2. SAR, Section 11.22
3. Ouid Cities 1 Nuclear Power Station first reloaa license submittal, Section 6.2.4.2, February 1974.

4 GE Topical Report NED0-20693, General Electric Boiling Water Reactor No. 1 Licensing submittal for Quad Cities Nuclear Power Station Unit 2, December 1974. Approvea revision number at time reload analyses are performed. 1.2/2.2-2a Amendment No. 83

1 O CP. OUAD-CITIES DPR-29 sidered inoperable. f ully provide reasonable assurance inserted into the core, that proper control red drive and electrically disarmed. performance is being maintained. The results of seasurements performed on the

5. If the overall average control rod drives shall be of the 20s insertion scram submitted in the annual operating time data generated to report to the NRC.

date in the current cycle esceeds 0.73 seconds, the MCPR operating limit must 3.The cycle cumulative mean be modified as required by scram time for 20s insertion Specification 3.5.K. will be determined tonediately f ollowing the testing required in Specifications 4.3.C.1 and 4.3.C.2 and the MCPR operating limit adjusted, if necessary, as reoutred by Specification 3.5.K. D. Castrel Red Aeroemleters D. Cearrel Red Aeromelesers At af. reactor operating pressum. a rod scru-Once a shift. cheet the status of the pmsure pulsior may be snoperable provided that no sad level alarms for each accumulator. other control rod in the nine rod square array around tha rod has

1. an inoperable scrumulator.
2. a direcuonal control vetve.iectncally daarmed whale ut a nonfully snarned possuon, of
3. a scram inneruofi greater than maa-unum prmissable annertson ume lf a control red with an inoperable acrumulator as annened fullin and its direcuonal control vaPves are elecincally disarmed, at shall not be considered to have an snoperable azumulator, and the rnd block staactated with thal anopera-ble earnulator may be bypassed L ReactMry a-lees L Rosethit) A-lies I

The rescuvity equivalent of the difference Durtng the stanup ust propam and stanups between the actual entacal rod cenagurauon following refueling outages. the entical rod and the espeoed conAgurauon during power sonagurauons wali be compared to the tspected oprauon shall not etceed Itik if this lams na con Agurauons at selened opersung condsaons escteded. the reaner shall be shuidown unut Thew cornparanons will be uwd as base data for the cause has been determined and correcuve reacuvity monitoring danng subsequent power actions have been taken in accordance with operauon throughout the fuel c)tle At speciac SynAcauen 6 6. the NRC shall be nouAed of po.et operaung conditions. the enucal rod this reponable occastrecs wtLhm 24 hours. een6gurauon will be compared to the coeng-arauon espected bawd upon apprepnatety cor-rened past data Tha compannon will be made at least every equivalent full power month F. Economic Genersties Castrol Symem F. Eremo=6e Generstles Camerol System Operation of the unit eith the economic gener. The rann set snio the economic generauon suon contro! sysiem with automatic flow con-control system shall be recorded weekly stol shal! be permissib4 only in the range of 65% to 1o01 af rated core now, with reactor power above 20% o i N. wos Amendment No. 83

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"W ^ '. QUAD CITIES 1.tra-29 C.- Scram Insertion Times O (j'.; The control rod system is analyzed to bring the reactor suberitical at a rate fast enough to prevent fuel. damage, i.e., to Drevent the MOPR from becoming less than the fuel cladding integrity safety limit. Xnalysis E the 1imiting 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 the requ. ired protection, and NCFR re=ains greater than the fuel cladding integrity 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 ODYU transient analysis computer code. The basis for choosing 0 73 seconds is discussed further i 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 ooserved time delay of about 210 milliseconds. Approxi=ately'90 millisec:nds 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. 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 inserti:n 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 wi!! be scrim tested , following a sh utdown. Scram times of new drives are approximately 2.5 to 3 seconds; lower rates of change in scram times following initis! plant operati:n at power are expected. The test schedule provides reasonable assarance of detection of slow drives belbre system deterioration beyond the limits of Specificanon 3.3.C.The program was developed on the basis of the sististical approach outlined below ( ar.d jud; ment. The history of drive performance accumuisted to daic indicates that the 90% insertion times of new and overhauled drives approximate a normal distribution about the mean which tends to become skewed toward longer scram times as operating timeis accumulated. The probability of a drive not exceeding the mean 90% insertion time by 0.75 seconds is greater than 0.999 for a normal distribution. The measurement of the scram performance of the drives surrounding a drive exceeding the expected range of scram performance will de:cet local variatier.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 possible anomslous performance. The numerical values assigncJ to the predicted scram performance are based on the analysis of the Dresden 2 startup data and of data from other HWR*: such as Nine Mile Point and Oyster Creek. {* The c<eurtence of ser.un times within the limits. but si;:nifcantly longer than sverage, should b: viewed as an indication of a syuematic pruhlem with conno! soJ Jrnes, especia!Iy if the number of drives exhibiting such scram tirnes steceds eight. the s!!owable number ofinoperahic rods. 3.3/4.3-10 Amendment No. $J ' U

.... - ~... :.u.. :. ....--.n....'.'-..-- - L' M QUAD CITIES DPR-29 within the prescribed limits within 2 hours, the reactor shall be brought to the cold shutdown condition within 36 hours. Surveillance and corresponding action shall continue until reactor operation is within the prescribed limits. Maximum allowable LHGR for all 8X8 fuel types is 13.4 KW/ft. K. Minimum Critical Power Ratio (MCPR) K. Minimum Critical Power Ratio (MCPR) During steady-state operation at The MCPR shall be determined rated core flow, MCPR shall be daily during steady-state greater than or equal to: power operation above 25% of rated thermal power. 1.39 (P8X8R) 1.37 (8X8/8X8R) forif 73 secs ave 1.44 (P8X8R) 1.42 (8X8/8X8R) for C,yg g.86 secs .385 27 + 1.109 (P8X8R) .385g((+1.089 (8X8/8X8R) for 734 E d.86 secs ave where2hve=mean20% scram insertion time for all surveillance data from specification 4.3.C whi'ch has been generated in the current cycle. For core flows other than rated, these nominal values of MCPR shall be increased by a factor of k where k is as shown in Figure 3.5.2. Ikanytimbduringoperation 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-state MCPR is not returned to within the pre-scribed limits within 2 hours, the reactor shall be brought to the cold shutdown condition within 36 hours. Surveillance and corresponding action shall continue until reactor operation is within the prescribed limits. Amendment No. f1, 83 3.5/4.5-10 w- .n n-

1 .............o_.. QUAD CITIES DPR-29 H. Condensate Pump Room Flood Protection See Specification 3.5.H I. Average Planar LHGR This specification assures that the peak cladding termperature following the postulated design-basis loss-of-coolant accident a will not exceed the 2200 F limit specified in the 10 CFR 50, Appendix K considering the postulated effects of fuel pellet densification. The peak cladding temperature following a postulated loss-of-coolant 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 red-to-rod power distribution within an assembly. Since expected local

  • variations in power distribution within a fuel assembly affect the calculated peak cladding temperature by less thant 20 F relative to the peak temperature for a typical fuel design, the limit on the average planar LHGR is sufficient to assure that calculated temperatures are below the limit.

The maximum average planar LHGR's shown in Figure 3.5-1 are based on calcu-lations employing the models described in Reference 2. Amendment No. 83

. e. ._.1 .c.....:.---w~.---- ,*. ! ' %i&M *k QUAD CITIES DPR.29 J. Local LHGR This specification assures that the maximum linear heat-generation rate in any rod is less than the design linear heat-generation rate even if fuel pellet densification is postulated. The power spike penalty is discussed in Reference 2 and assumes a linearly increasing variation in saial gaps between core bottom and top and assues with 955 confidence that no more than one f uel roc exceeds the design LHGR oue to power spiking. No penalty is required in Speciftcation 3.5.L because it has been accounted for in the reload transt.ent analyses by increasing the calculated peak LHGR by 2.25. K. Minimum Critical Power Ratio (MCPR) The steady state values for MCPR specified in this specification were selectea to provice margin to accomodate transients and uncertainties in monitoring the core operating state as well as uncertainties in the critical power correlation itself. These values also assure that operation will be such that the irtittal condition assumed f or the LOC A analysts plus two percent for uncertainity is satisfiec. For any of the special set of transients or disturoances caused by single operator error or single equipment malfunction, it is required that design analyses intttall2eo at this steady-state operating limit yield a mCP3 of not less than that specifted in Specification 1.1.A at any time curing tee transtent, assuming instrument trip settings given in Specification 2.1. For analysis of the thermal consequences of these transients, the value of MCPR statec in this specification f or tMe limiting concition of operation Downcs tne initial value of MCPR assumec to exist prior to the inittation of the transier:5. This initial condition, which is usec in the transient anaiyses, will preclude violattor of the fuel ciaoding integrity safety limit. Assumptions anc methods usea in calculating the requireo steacy state MCPR limit for each reloso cycle are documented in References 2, 4, enc 5. The l results apply with increased conservatism while operating with MCPR's greater than specified. The most limiting transients with respect to MCPR are generally:

4) Roo withdrawal error b) Load rejection or turbine trip utthout bypass c) Loss of feedwater heater Several factors influence which of the these transients results in tne largest reduction in critical power ratto such as tne spectftc fuel loading, encosure, anc fuel type.

The current cycle's reloac licensing analyses spectftes the limiting transients for a given esposure increment for eacn fuel type. Tne values specifies as the Ltmiting Condition of Operation are conservatively chosen to bound the most restrictive over the entire cycle for each fuel type. The neeo to adjust the MCPR operating limit as a function of scram time arises from the statistical approach used in the implementation of the ODYN computer code for analy2ing rapid pressurization events. Geneetc stattstical analyses were performeo for plant groupings of sistlar design ehten considered the statistical vartation in several parameters (initial power level, CRD scram, insertion time, and model uncertainty). These analyses (which are oescrtDeo further in Reference 4) produced generic Statistical adjustment Factors which have been applied to plant ano cycle specif ic 00YN results to yield operating limits which provice a 951 pecca 111ty with 951 conftdence that the Ilmiting pressurization event =til not cause MCPR to fall below the f uel clacctng Integrity safety Itatt. 3.5/4.5 14 Amendment No. g7, 83

O -....,-..........+...-....:. .. : : ;. w.w t QUAD-CITIES 'N DPR-29 As a result of this 95/95 approach, the average 20% insertion scram time must be monitored 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 compared to the distribution assumed in the ODYN 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 appliea to the plant specific ODYN results (i.e. without statistical adjustment) for the limiting single failure pressurization event occuring at the limiting point in the cycle. It is not applied in full until the mean of all current cycle 20% scram times reaches the 0.90 secs value of Specification 3.3.3.C.l. 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 requirea 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: 7~8 = 4 + 1.65 (Nj/ Nj)l/2 o-

-i where Jf = mean value for statistical scram time distribution to 20% inserted cr = standard deviation of above distribution N1 = number of rods tested at.80C (all n

operable rods) 7_N i i = total number of operable rods tested in the current cycle T h e v a l u e f o r 7's u s e d 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 j 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 l months. That is, a cycle length was assumed which is longer l than any past or contemplated refueling interval and the number of rods tested was maximizeo in order to simplify and conservatively reduce the criteria for the scram time at which l MCPR penalization is necessary. b) The values of Wand C" 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 Specification 3.5.k initiates the linear MCPR penalty at a slightly lower v a l u e 7"a v e. This also produces the full 4.4% penalty at 0.86 secs which would occur sooner than the requried value of 0.90 secs. 3.5/4.5-14a Amendment No. 83

a- -2=...-...-.o- .L.. =*M :: a QUAD CITIES ' DPR-29 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 speed controller causes the scoop tube positoner 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," NED0-24146A*, April, 1979 2. " Generic Reload Fuel Application," NEDE-240ll-P-A** 3. I. M. Jacobs and P. W. Marriott, GE Topical Report APED 5736, " Guidelines for Determining Safe Test Intervals and Repair Times for Engineerea Safeguards," April, 1969. 4. " 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-24154 Vol. III as supplementea by letter datec 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". Approved revision at time of plant operation. Approved revision number at time reload fuel analyses are performea. 3.5/4.5-15 Amendment No. H, 83

.~, + QUAD-CITIES DPR-29 b5 P"

2. Both the sump and air sampling sys-sems shall be operable during reactor power operation. From and after the date that one of these systems is made or found to be inoperable for any rea-son, reactor power operadon is per-missible only during the suczneding 7 days.
3. If the conditions in i or 2 above can-not be met an orderly shutdown shall be initiated and the reactor shall be in

-a cold shutdown condition within 24 hours. E. Safer) and Relief Yahes E. Safety and Relief Yahes

1. Prior to reactor startup for power op-A minimum of I/2 of all safety valves shall be cration, during reactor power operat-bench checked or replaced with a bench ing conditions, and wl,enever the reac-checked valve each refueling outage. The pop-tot coo! ant pressure is greater than 90 ping point of the safety valves shall be set as psig and temperature greater than follows:

320' F. all nine of the safety valves sha?l be operable. The solenoid-Number of VaMs Serpoint (psig) activated pressure valves shall be oper-1 1135ni l able as required by Specification 2 1240 2 1250 2. If Specification 3.6.E.1 is not met, the 4 1260 reactor shall remain shut down until the condition is corrected or, if in The allowable setpoint error for each valve is operation, an orderly shutdown sha!!

  • 1%

be initiated and the reactor coolant All relier valves shall be checked for set pres-5"Ik

  • f sure each refueling outage. The set pressures p

hours. Number of Valves Serpoint (psig) I s 1135"' i 2 sII15 2 51135 { "rTarges Rock combination safety / relief valve. F. Structura! Integrity F. Structurallategrity The structural integnty of the primary rystem The nondestructive inspections tieted in Table boundary shad be mamtamed at the lev:1 re. 4.61 shad be performed as specified in accor-quired by the ASME Boiler and Pressure %ssel dance with Section XI of the ASME Bocer and Code, Section XI, " Rules for Innervice Inspection Pressure Vesse! Code,1971 Edition, Summer of Nuclear Power Plant Components",1974 1971 Addenda. The results obtained from com-Edation Summer 1975 Addenda (ASME Code phance with this specification will be evaluated Section XI). O after 5 years and the conduzions wiD be reviewed L, with the NRC. 3.6 / 46-d Amendment 33

  • Amendment No. 83

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