ML20009H267
| ML20009H267 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 07/27/1981 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML17194A064 | List: |
| References | |
| NUDOCS 8108070123 | |
| Download: ML20009H267 (20) | |
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Attachment 1 Quad Cities Station Unit 2 Proposed Changes to Appendix A, Technical Specifications to Facility Operating License DPR-30 Revised Pages: 1.1/2.1-4 1.1/2.1-5 1.1/2.1-6 1.1/2.1-7 1.1/2.1-11 1.2/2.2-1 1.2/2.2-2 3.3/4.3-5 3.3/4.3-10 Figure 3.5-1 ( Sheet 1 of 5)
Figure 3.5-1 (Sheet 2 of 5)
Figure 3.5-1 (Sheet 4 of 5) 3.5/4.5-10 3.5/4.5-14 3.5/4.5-14a New Pages: 1.1/2.1-7a 1.2/2.2-2a 3.5/4.5-14b
. 0108070123 810727 PDR ADOCK 05C70265-P PDR
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QUAD-CITIES DPR-30 1.1 SAWTY LIMIT 19S13 The f'uel cladding integrity linit is set such that no calculated fuel drewje would occur as a resu1* of en abnormal operational tr.nsie nt. Decause fuel damage is not tirectly ebeca vabic, a situp back atta ocel. is used to establich a safety linit auch that the minimuss critical power ratio (ncrat) in no jesa than ne fus t cladding integrity safety linit.rCrn > the fuel cladding integrity safety linit representa a cons,ervative margin reistive to the conditaons required to maintain fuel cladding integrity.
The fual cladding to c,no of the physical buundaries which separa te radioactivu materials from the cravirons, The integrity of the fuel cladding la related to its relative frevdam frun performations or cracking.
Although some corrosion or use-related cracting osy occur during the life c.f the cladding, fissien product migration from this source is inerenen* ally cur-J1ative and continuounty ceakuteble. fuel claddang per.
forations, however, can result from therr.cl stas recs which occur fro.e s cactor operatiets significant1v alove design conditions and the protection sytten safety settings. lih11w fisoson product afgratien fres elsddir:;
perforation is juu as eieasu.able a,s that frem u*c-related crackirug, the therually c used claddisy pertos.
ations signal a threshold Ltyond shach still greater thermal stresses may cause gaou ratPer than ir.crt ent.
al cladding deterioration. Therefore, the fuel eleddtng anfety limit is duf tned with snargin to the con * .
tions thich would produce onset of transt t aen bos1tng (nCPA of 1.0). These conditions represent a signiti-cent departure fros the condition intended by design for ptshned operation. Therefore, the fuel clerMang integrity safety lirait is Established such that no calculated fuel dmage Sh311 result from an abnornal operational transient Baats of the values derived for thi. r.afety limit for each fuel type is documented in Reference 1 Rud 2. g A. Reactor pressure > 800 psig and Core Flow > 1C1C of Stated Onset of tronettion 1,o111ng results in a decrease in heat trarafer froat the claddisig and therefore
. elevated claddirg tersperrture and the posnacility of claddir.g failur e. flowever, the existe nce of critical poeer, or boiling transition is not a directly observablu g artmeter in an opsratirwf react-or. Therefoto, the margin to boiling trana.ition is em1culated fresa plent opcreting parrmetra s such ae core power, core floe, fee 6eter serveratura, and core power dirte sbut et.n. TP.e margin for es e?.
fuel asser*1y is characterized by the critical power ratio tirk). shich as tha ratio of the i.aMie
- - power which would p educe entiet cf transition boiling divided I.,y the actual bue.dle power. 11.e minimura value of this ratio for cry bundic in the core is the minimum critacol power ratio (Mrn).
It 's essumed that the pler.t orcratien is controlled to the norninal protecttve setponta via the instrumented variables (rigure 2.1-3).
Tho #1CPR fuel cladding integrity safety limit has suf ficient con;crvatisms to assure that in the everc.
of an ebnorw.a1 operats oul trensicei'. init1hted train the norruil opci at ang condition, s.orr tenn 99,p.;
of the fuel rods in the core are capected to avoid boiling transition. The margin bethuen hCPR of 1.0 (onset of transition boiling) and the. mafety linit, is derived frc..i a detailed statistical analysim considering all of the uncertaintics in sionatoring the core operating state. includarg uncertainty in the boiling transition corrulation (see e.g., keferenec 1). Because the bo 113.g transition correlation is based on a large quantity of full-scale stata3 the re is a very high con.
fidence that operation of a fuel assenhly at the conditio.1 of 71 CPM = the it.e1 cladding integr.ty safety Ibnit would not produce boiling transition.
However, if boiling transition were to occur, cladding perforation would not be expected. C3 rddire temperatures would increanc to appro::iantely 1100'r, which la lerlew the perf oration temperntse e of the cladt'ing rsaterj al. This has been verified by teste in the Ceneral 1:1cct ric Test Reactor (Cr'fr.) ,
where similar fuel operated ebove the critical heat flux for a signi!Jeant period of time (30 r.iin-utes) withmat cladding parforation.
If reactor pressure should ever exceed 1400 psia during normal power operatlon (the limit of applicability of th<r boiling transition correlatien), it would be assumed that the fLc1 claddtog intec ity safety limit has been violated.
In addition to the boiling tranettien limit trCPR, operation la constrained to a smaximum Ls4Cas17.5 kw/f t for 7 x 7 fuel arul 13.4hw/f t for all 8x8 fuel types. This constraint is established by specification 3.3.3.
strain for abncrmal to nrovide adequ:te operating safety margin transients to, ated ini: rom nigh1ts
. plastic power conditions. Specification 2.1. A.1 provides for equivalent safety margin for transients initiated from lower power con-ditions by adjusting the APEM flow-biased scram setting by the ratio of FRP/MFLPD. .
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QtlAD.Cilll5 DPR .10 Speci5 cation 3.5) established the LilGR maximum which cannot be exceeded under steady power operation.
B. Core Thermal Power Limit (Reactor Pressurc<R00 psia)
At pressures below 500 psia, the core elevation pressure drop (0 power,0 flow)is greater than 4.56 psi.
At low powers and Dows this pressure differentialis maintained in the bypass regien of the core. Since the pressure drc p 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 now of 28 x 10'lb/br bundle now, bundle pressure drop is nearly independent of bundle powe: nd has a value of 3.5 psi.'Ihus ti. bundle flow with a 4.56-psi driving head will be greater than 28 x 10'Ib/hr. Full seale ATLAS test data taken at pressures from 14.7 psia to SCO psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. At 25% of rated thermal power. the peak powered bundle would have to be operating at 3.86 times the a<erage powered bundle in o. der to achieve this bundle power. Thus, a core thermal power limit cf 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 neutron !!ux due to the inherent heat transfer time constant of the fuel, which is 8 to 9 secondt Also. the limiting safety system scram settings are at values which will not allow the reactor to be operated above the safety limi
/h during normal operaticn or during other pla.it oper.iting situations w hich have been an lyzed in detail.
In addition, control rod scrams are such that for normal operating transients. t'ae neutron flux transieni is terminated before a significant increase in surface heat fiux 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.
Exceeding a neutron Dux scram setting and a failure of the control rods to reduce flux to less than the scram setting wtthin I.5 seconds does not necessarily imply that fuelis damaged; however. for this specification, a safety limit viol: tion will be assumed any time a neutron Dux scram setting is exceeded for longer than 1.5 seconds.
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 operatir.g trarsients expected.These analyses show that even if the bypass system l fails to operate, the design limit of MCPR = the fuel cladding intearity safety l
limit is not exceeded. Thus , use of a 1.5 second limit provides in.
additional The computer mbr8has provi a sequence annunciation prograrn which will hdicate the sequence in which scrams occur,;uch as neutrcn flux, pressure.cic This program also indicates when the scram setpcint is cicated. This will previde information on how loag a scram condition exists and thus provide some rneasure of the energy added during a transient. Thus, computer information normally will be available l
I for analyzing scrams; however, if the computer information should not he available for any scram
! analysis, Specitication 1.1 C.2 will be relied on to determine if a safciy limit has been violated.
During periods when the reactor is shut dowr. consideration must also be given to water level reovirements due to the effect of decay heat. If reactor water level should drop below the top of the active fuel during this time, the abdity ta ecol the core is reduced. This reduction in core-cooling cap.M!ity could lead to elevated claddmp temperatures and ciaodmg perforation.The score will he cooled wiric.ently l to prevent cladding melting should the water lesel be reduced,to two. thuds the nue heiEht 1.si.shhsh.
I i ment of the ufety limit at 12 ir.ches above the top of the fuel provides adesgu.ne in.irpn. 'lhis level will be continuously monnared whenever the rnisculanon pumps are not operating.
- Tcp of the active fuel is defined to be 360 inches above vessel zero (see Bases 3.2).
- 1. l / 2.1 - $
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, References .
- 1. " Generic Reload fuel Applications," NEDE-24011-P-A*
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- 2. " Generic Informa tion For Barrier Fuel Deaonstra tion Bundle Licensing", NElX)-24259-A, February 1981. ,
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- Approved revision number at time reload fu i anal iy es are performed.
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OPR-30
2.1 LIMITING SAFETY SYSTEM SETTING BASIS The abnormal operational transients applicable to operation of the units have beer, 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 ana 2. Transient analyses are initiated at the conditions given in these References. l I
The scram delay t irte and rate of rod insertion allowed by the analyses are conservatively set equal to the longest delay and I slowest insertion rate accepteble 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% insert ~ ion. 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 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.
l Steady-state operation without forced recirculation will not be permittea except during startup testing. .The analysis to support operation at various power and flow relationships has considered operation with either one or two recirculation pumps.
The bases for individual trip settings are discussed in the following paragraphs.
For analyses of the thermal consequences of the transients, the MCPR 's stated in Paragraph 3.5.K as the limiting condition of operation bound those which are 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 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 tuel.
1.1/2.1-7
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Therefore, during abnormal operational' transients .the thermal power of the fuel will'be less than that indicated 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.
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QUAI)-CITil'.S DPR-30 4
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- References
! 1. " Generic Reload Fuel Application," ImDE-24311 2P-A*
- Approved revision number at thne coload analyses are perf armed
- 2. " Qualification of the One-Dimensional Core Trancient Model for Boiling Water Reactors" General Electric Co. Licensing Topical Report NEDO 24154 Vols. I and II and NbDE-211194 VolumeII .III au supplemented by letter dated deptember 5,1980 from R.
Buchholr. (GE) to P. S. Check (NRC).
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QllAD-CITIES DPR-30 1.2/2.1 REACTOR COOLANT 5YSTEM i
LIMITING SAFETY SYSTEM SETTING SAFETY LIMIT Applicabillry:
Applicability:
Applies to trip settings of the instruments and Applies to limits on r :etor coolant system devices which are provided to prevent the reactor pressure.
system safety limits from being exceeded.
Objective:
Objectise:
To defme the 1 vel of the process variables at which To establish a limit below which the integrity of the automatic protective action is initiated to prevent reactor coolant system is not thseatened due to an the safety limits from being execeded.
overpressure condition.
SPECIFICATIONS 6
um A. Reactor coolmt high pressure scram shall be gr ... .. . ..ru 6,
- a. w. re-tor e-a.nt .,.t..
' '" s1060 psig.
I'!;**.Il*:I 3 7Ei%'. %f,*t'!O!:t% C I'i. ;r %%"
in the re.ctor vesses B. Primary system safety valve nominal setpags shall be as follows:
I valve at lll5psig
2 valves at 1240 psig 2 valves at 1250 psig 4 valves at 1260 psig tirTarget Rock combinatinn safety / relief vahe The allowable setpoint error for each valve
- shall be i1%.
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QUAD CITIES DPR-30 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 5600F. The pressure 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 B31.1 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 evaluaton 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 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 documented 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 designed for a general membrane stress no greater than 26,700 psi at an internal pressure of 1250 psig; this is a i factor of 1.5 below the yield strenght of 40,100 psi at 5750F. At the pressure limit of 1375 psig, the general l membrane stress will only be 29,400 psi, still safely below the
( yield strength.
The -elationships 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.
l The normal operating pressure reactor coolant system is.1000 psig. For the 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 addition, l pressure relief valves have been_provided to reduce the
! probability of the safety valves operating in the event that the
. turbine bypass should fail.
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QUAD CITIES DPR-30 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, nowever. 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
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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 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.
- Approved revision number at time reload analyses are performed.
1.2/2.2-2a
.' s QUAD-CITIES DPR-30 sidered inoperable, fully provide reasonable assurance inserted into the core, that proper control rod drive and electrically disarmed. performance is'being maintained. The results of measurements performed on the
- 5. If the overall average control rod drives shall bt of the 20% insertion scram submitted in the annual operating time data generated to report to the NRC.
date in the current cycle exceeds 0.73 seconds, the MCPR operating limit must 5. The cycle cumulative mean be modified as required by scram time for 20% insertion-Specification 3.5.K. will be determined immediately following the testing required in Spec if ications 4.3.C.1 and 4.3.C.2 and the MCPR operating limit adjusted, if necessary -as required by 50ecification 3.5.K.
D. Control Rod Accumulators D. Contral Rod Accumulators At all reactor operating pressures, a rod accu- Once a shift. check the status of the preuure mulator may be inoperable provided that no and level alarms for each accumulator other control rod in the nine-rod square array around thas rod har
- 1. an inoperable accumulator.
- 2. a directional control valve electrically disarmed shale in a nonfully inserted position. or
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- 3. a scram insertion greater than man-imum permissible insertion time.
If a control rod with an inoperable accumulator is inserted full-in and its directional control valves are electrically dnarmed. :: shall not be considered to have an inoperable accumulator, and the rod block associated with th.at inopera- ,
ble accumulator may be bypaswd. 1 E. Reactivity Anomahes E. Reactisity Anomatics The reactivity equivalent of the difference During the startup test program and startups between the actual critical rod configuration following refueling outages, the critical rod and the expected con 6guration during power configurations will be compared to the expected operation shall not exceed ILik. If this hmit is configurations at selected operating conditions.
exceeded, the reactor shall be shutdown until Thew comparisons wi ll be used as base data for the cause has been determined and corrective reactivity monitoring during subsequent power actions hase been taken. In accordance with operation throughout the fuel cycle. At specihc Spec 6 cation 6 6. the NRC shall be notined of power operating cre ditions, the critical rod i th2s reportable occurrence within 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 config-uration expected based upon appropriately cor-rected past data This comparison will be made at least every equivalent full power month.
F. Economic Generation Control System - F. Economic Generation Control Sprem Operation of the unit with tre economic gener- The range set into the economic generation
- ation control system with automatic now con- control system shall be recorded weekly.
trol shall be permissible only in the range of 65% to 100% of rated core now, with reactor power above 20%.
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QUAD CITIES s urn-bo I .C. Scram Insertion Times The control rod system is analyzed to bring the reactor suberitical at
], a rate fast enough to prevent fuel damage, i.e., to prevent the MCPR from becoming less than the fuel cladding inte6rity safety limit.
~nnalysis E the' limitin9 Power transient shows thrat the negative reactivity rates resulting from the scram with the average response of all the drives as given in the above specification, provide the required protection, and MCFR remains 6reater than the fuel cladding inteErity safety limit. It 1 necessary to raise the MCPR operating limit (per Specification 3.5.K) when the average 20fs 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 further in the bases for Specificatior: 3.S.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 campared 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 insertiun 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 scrim 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 deterioration beyond the limits of Specification 3.3.C. The program was developed on the basis of the statistical approach outlined below
( and judgment. -
The history ofdrive 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 toward longer . scram ilmes as operating time is accumulated.The probability of a drive not exceeding the mean 90% insertion time by 0.75 seconds is greater than 0.999 for a normal distribution. The measurement of the scram performance of the drives surrounding a drive exceeding the expected range of scram perfortnance will detect lodal 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 possib!c 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 HWR's such as Nine Mile Point and Oyster Creek.
The occurrence of scram times within the limits. but si; nificantly longer than average. should b: viewed
(' ' as an indication of a systematic problem with contml rod drives, especially if the number of drives exhibiting such scram times cacceds eight, the allowable number ofInoperabie rods.
3.3M3-10 1
DPR-30 OUAD CITIES UNIT 2 '
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9.5 ' ,
10;000 20,000 30,o'o'o 40,000 Planar Avarage Exposure (MWD /ST)
Figure 3.5-1 Maximum Average Planar Linear Heat Generation (Sheet 4 ot' 5) Rate (MAPLHCR) va. Planar Average Expsoure
- ,-
i- *
- -QUAD CITIES 1
.DPH-30 '
'within the prescribed brnits 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 cc i responding action shdi continue until rea. tor
, "Peration is withia the prescribed brnits.
i Maximum allowable LHGR for all-
~ ~ ~ _ 8XG fuel types is 13 4 KW/ft.
I K. Minimum Critical Power K. Minimum Critical Power
During s teady-sta te opera tion The MCPR shall be determined g 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.
- I 1.37 for Tave 5 0.73 secs 1.42 for Tave 2 0.86 ' sees o.385 Tave + .1.089
- for. o.73 < Tave < o.86 secs T
where Tave = mean 20% scram 1 insertion time for all surveillance data from Specification 4.3.C. which has been t
generated in the
- current cycle.
I For core flows other than rated,
-these nominal values of MCPR shall' i- .be increased'by a factor of kg
-where kr is as shown in Figure 3 5.2.
3 If any time during operation 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-i sta te MCPR is not returned to within i 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 /> i
-the reactor shall be brought to the ,,,
t cold' shutdown Surveillancecondition and within 36 correspond-
. hours.
ing action shall continue until l
reactor operation is within the prescribed limits.
!- 3 5/4.5-10 t
N
QUAD CITIES DPP-30
', shown on Figure 3.5-1 as limits because conformance calculations have not been perf ormed to justify operation at LHGR's in excess of those shown.
J. Local LHGR Tris 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 penalJy is discussed in Neference 2 and assumes a linearly increasing vo-iation in axial gaps between core bottom and top and assues with 951 confidence that no more than one fuel rod exceeds the design LHGR due to power spiking. 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 LHGR by 2.21.
K. Minimum Critical Power Ratio (MCPR)
The steady state values for MCPR specified in this specification were selected to provide 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 intitial condition assumed for the LOCA analysis plus two percent for uncertainity is satisfied. For any of the special set of transients or disturbances 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 transient, 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 condition of operation bounds the iritial value of MCPR assumed 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 cladding integrity saf ety limit. Assumptions and methods used in calculating the required steady state MCPR limit for each reload cycle are documented in References 2, 4, and 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:
a) Rod withdrawal error b) Load rejection or turbine trip without bynass l c) Loss of feedwater he**er Several factors influence which of the these transients results in
, the largest reduction in critical power ratio such as the specific l fuel loading, exposure, and fuel type. The current cycle's reload licensing analyses specifies the limiting transients for a given exposure increment for eact fuel type. The values specified as the i
i
^
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 f unction of scram time arises from the statistical approach used in the implementation of the ODYN computer code for analy2ing rapid pressurization-events. Generic statistical analyses were perf ormed f or plant groupings of similar design which considered the statistical variation in several parameters (initial power level, CRD scram insertion time, and model uncertainty). These analyses (which are cescribed further in Reference 4) produced generic Statistical i Adjustment Factors which have been applied to plant and cycle i specific ODYN results to yield operating limits which provide a 951 probability with 95% confidence that the limiting pressurization event will not cause M;PR to f all below the f uel cladding integrity safety limit.
3.5/4.5-14
(
QUAD-CITIES OPR-30 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 applied 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 required 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:
T=4+ 1.65(N/fNj)l/2 B 1
- =i 7
where Af = mean value for statistical scram time k CF distribution to 20% inserted I
= standard deviation of above distribution N 1 = number of rods tested at BOC (all g operable rods)
AN; = total number of operable rods tested in I
the current cycle The v a l ue f or 7'B u sed in Specification 3.5.k is 0.73 secs which is conservative for 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 months. 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 CT 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 d slightly l o we r v a l u e 7"a v e . This also produces th' Tull 4.4%
penalty at 0.86 secs which would occur sooner than the requried v.lue of 0.90 secs.
- 3. 5 /4. 5 - 14 a
' 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 j 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," NE90-24146A*,
April, 1979
- 2. " Generic Reload Fuel Applica'. ion," 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 Engineered 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 supplemented by letter dated September 5, 1980 from R. H.
Buchholz (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.
- 1 Approved revision number at time reload f uel analyses are performed.
- - 3.5/4.5-14b f
i 5
s> .
a l
Attachment 2 ERRATA AND ADDENDA SHEET NO. S to NEDD-24146A ERRATA AND ADDENDA SHEET NO. 6 to NEDD-24146A
_. ._ - -