ML20063C433
| ML20063C433 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities, 05000000 |
| Issue date: | 08/19/1982 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML20063C409 | List: |
| References | |
| NUDOCS 8208270256 | |
| Download: ML20063C433 (20) | |
Text
.
Quad Cities Station Unit 1 Proposed Changes to Appendix A, Technical Specifications to Facility Operating License DPR-29 Revised Pages:
1.1/2.1-4 1.1/2.1-5 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 3.5/4.5-10 3.5/4.5-14 3.5/4.5-15 Figure 3.5-1 ( Shee t 1 o f 4)
Figure 3.5-1 (Sheet 2 o f 4)
Figure 3.5-1 (Shee t 3 o f 4)
Figure 3.5-1 (Sheet 4 o f 4)
New Pages :
1.1/2.1-7a
- 1. 2 /2. 2 -2a 3.5/4.5-13a 3.5/4.5-14a 4658N 8208270256 820919 PDR ADOCK 05000254 p
(7 i
\\
OUAD-CITIES DPR-29 1.1 SAWTY LIMIT 13 ASIS The [uel cladding integrity linit is set such that no enlculated fuel dwonge would occur as a resu1* nf en abneraal operational trinsient. Uccause fuel damage is not directly observabic, a step-back apps ocet. is used to establish a safety limit such that the minimum critical power ratio (MCPU) is no 3 cts than the fusi cladding integrity safety linit r. Crit > the fuel cledding integrity safety limit representa a cont,civative margin relative to the conditions required to maint win f uct cladding integrity.
Tne fuel cladding la one of the physical boundaries which aparate radioactive materials from the environs.
The integrity of the fuel cladding is related to its relative freedam frun performations or crackin6-l Although some corrosion or use-related cracking may occur during the life of the cladding, fission product migration from this source is incrementally ct.r ulative and continuously censureb1c. Fuel c1cddang per-forations, however can result from thernc1 atterzes which orcur fro:s scoctor operatien significant1v above design conditions and the protection system safety settings. While fissaan product migratien fre's c'IMd a ny perforation is ju*.t ts sneasurabic as that f rem uec-related cracking, the therually c.iused claddia.9 perios.
ations signal a threshold beyond whach still greater thermal stresses may cause gaohs rather than inert.vnt-al cladding deterioration. Therefore, the fuel cledding safety limit is defined with raargin to the cenja-tions which would produce onset of transit aon boaling (MCPR of 1.0).
These conditions represent a e s qniti-cant departure froa the condition intended by design for planned operation. Therefore, the fuel cler integrity safety limit is established such that no calculated fuel demage shall result from an?':ng abnornal operationni transient. Basis of the values derived for this safety limit for each fuel type is documented in Reference 1 A.
Reactor Pressure 7 800 psig and core Flow /10% of Rated Onset of transition boiling results in a decrease in heat transfer fro i the claddirig and thernfore elevated cladding temperrture and the possi' ility of cladding f ailur e.
However, the existence of c
critical power, or boiling transition is not a directly cbservablo paremeter in an operatante re.ict-or. Therefore, the raargin to boilang trene.ition is calculated f reu plent operating parrmetes s such as core power, core flov, feedwater tenperature, and core power distr ibut ion.
The margin for eses fuel assembly is characterized by the critical power ratio (CPR), which as the ratio of the bandle power which would prcduce onset of transition boiling divided by the cetual buhdle power. The minimum value of this ratio for eny Intndle in the core is the minimum critical pm.er ratio (MCPR).
2L is escumed that the plant operatien is controlled to the nominal protect 1ve r.rtponts via the instrumented variables (Figure 2.1-3).
The 91CPR fuel cladding integrity sa'ety limit has suf ficient con;crvatism to asbute that in the cven.
of an ebnormal operr.tionc1 transient initiated frein the norval oper ating een.lition, a.or e tnan 99.?;
of the f uel rewis in the coac are expoeted to avoid boiling transition. The margin between !;CPR of 1.0 (onset of transition poiling) and the safety limit, is derived frca a detailed statistical analysim considering all of the uncertainties in monitoring the core operating state, includang uncertainty in the boiling trcnsition correlation (see e.g., hef e re nce 1), accause the bos1119 transition correlation is based on a large quantity of full-senic data, there is a very high con-fidence that operation of a fuel essembly at the condition of HCPR = the fuel cladding intcGr ty safety limit would not produce boiling transition.
However, if boiling transition were to occur, cladding perforation would not be expected. c1 coding l
temperatures would increar.c to appro::isntely 1100 f, which is below the perforation temperatric of I
the cladding raaterial. This has been verified by tests in the cenera) Electric Test Reactor (Crtn).
where similar fuel operated cbove the critical heat flux for a significant period of time (30 rain-utes) without cla,dding perforation.
If reactor pressure should ever execed 1400 psia during normal power operation (the limit of applicability of thu boiling transition correlation), it would be assused that the fuel cladding integrity safety limit has been violated.
In addition to the boiling transitien limit (MCPR) operation is constrained to a maximum LH CRs17.$
kw/ft for 7 x 7 fuel and 13.4hw/ft for all ex8 fuel types. This constraint is established by
- margin to 15 plastic strain for abnormalrovide adecuate Safetyoperating" transients init atec from nigh specification 3.b.J.
to "
j j
power conditions.
Specification 2.1.A.1 provides for equivalent safety margin for transients initiated from lower power con-ditions by adjusting the APM flow-biased scram setting by the ratio of FRP/MFLPD.
1.1/2.1-4
~
r QtlAI)-Cl t il:S
(
DPR-29 Speci5 cation 3.5J established the LIIGR maximum which cannot he exceeded under steady power operation.
B.
Core Thermal Power Limit (Reactor Pressurc<S00 psia)
At pressures bc!ow SJO psia, the core elevation pressure drop (G pow er, O flo v)is greater than 4.56 psi.
At low powers and flows this pressure differentialis maintained in the bypass region of the core. Since the pressure drop in the bypass region is essentially all elevation head. the core pressure drop at low powers and flows will always be greater than 4.56 psi. Analyses rhow that with a flow of 28 x 10'Ib/hr I
bundle flow, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi.'!hus the bundle flow with a 4.56 psi driving head will be greater than 28 x 10'Ib/hr. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. At 25% of rated thermal power. the peak powered bundle would have to be operating at 3.86 times the average powered bundle in 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 behind the neutron flux due to the inherent heat transfer time constant of the fuel.which is 8 to 9 secon$. Abo. the limiting safety h
system scram settings are at values which will not allow the reactor to he operated above the safety limit f's during normal operation or during other plant operating situations s hich have been analyzed in detail.
In addition, control rod scrams are such that for normal operating transients, the neutron flux transient is terminated before a significant increase in surface heat flux occurs. Control rod scram times are checked as required by Specification 4.3.C.
and the MCPR operating limit is modified as necessary per specification 3 5.K.
Exceeding a neutron flux scram setting and a failure of the control rods to reduce flux to less than
)
the scram setting within 1.5 seconds does not necessarily imply that fuel is damaged; however, for this i
specification, a safety limit violation will be assumed any time a neutron flux 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 operating transients expected. These analyses show that even if the bypass system fails to operate, the design limit of MCPR = the fuel cladding intearity safety i
limit is not exceeded.
Thus, use of a 1.5 second limit provides additional marain.
The computer providt d'has a sequence annunciation prograr6 which w.!! ind.icate the sequence m. which n
scrams occur, such as neutren flux, pressure, etc. This program also indicates when the scram setpeint is cleared. This will previde information on how long a scram condition exists and thus provide some rneasure of the energy added during a transient.Thus, computer information normally will be available for analyzing scrams; however, if the computer information should not he available for any scram j.
analysis, Specitication I.I.C.2 will be relied on to determine if a safety limit has been violated.
During periods when the reactor is shut down, consideration must also be given to water level requirements due to the effect of decay heat. If reactor water level should drop below the top of the active fuel during this time, the ability to cool the core is reduced. This reduction in core. cooling capability could lead to elevated cladding temperatures and cladding perforation.The core will he cooled wiriciently to prevent cladding melting should the water level be reduced,to two-thuds the cose height 1.stabbsh.
(
ment of the safety limit at 12 inches above the top of the fuel provides adequate margin.'l his level wdl be continuously monitored whenever the rrritculation pumps nre not operating.
l
- Top of the active fuel is defined to be 360 inches above vessel zero (see Bases 3.2).
1.1/ 2.1 -5
Quad Cities DPR-29 2.1 LIMITING SAFETY SYSTEM SETTING BASIS-(-s)
The abnormal operational transients applicable to operation of the units have been analyzed throughout the spectrum of planned s/
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 and 2.
Transient analyses are initiated at the l
conditions given in these References.
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 cod 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 steady-state condition.
D)
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 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 l
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
\\
\\s-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.
1.1/2.1-7 i
e
J k
1 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.-
i
'I d
i l
l 1
4 i
l t
4 i
i ll-i i
i' i
i I
l 1
i i
l r
i t
l 1.1/2.1-7a
,c__._
e e
I i
QUAD-CITIES
)
)
{
l t
t i
References i
1 1.
" Generic Reload Fuel Application," !!EDE-240ll-P-A*
- Approved revision number at time roload analyses are performed l
- 2. " Qualification of the One-Dimensional Core Transient Model for l
Boiling Water Reactors" General Electric Co. Licensing Topical l
Report NEDO 24154.Vols. I and II and NEDE-24154 Volume III as l
supplemented by letter dated September 5,1980 from R. H.
l Buchholz (GE) to P.
S. Check (NRC).
t I
+
4 I
e f
J I
f i
4 4
1.1/2.1-11 i
i i
a l
i
Qtf AD-CITIES
(,f m')
DPR-2 9 C/
1.2/2.1 REACTOR COOLANT SYSTEM LIMIT 1NG SAFETY SYSTEM SETTING SAFETY LIMIT Applicabillry:
Applicability:
Applies to trip settings of the instruments and Applies to limits on r::ctor coolant system devices which are provided to prevent the reactor pressure.
system safety limits from being exceeded.
Objectise:
Objectise:
To establish a limit below which the integrity of the To define the level of the process variables at which reactor coolant system is not thie.itened due to an automatic protective action is initiated to prevent the safety limits from being exceeded.
overpressure condition.
SPECIFICATIONS A.
Reactor coolant high pressure scram shall be a.
no re.cter co i.nt.,.te. vr....r......orea e, sne IW'l.II*:I 'O*IiC",l%"iEr".'dl'M 2*Fif'r;%;'
51060 psig.
in tr r..c t o r.... 1.
B.
Primary system safety valve nominal settings shall be as follows:
1 valve at 1115psig'"
2 valves at 1240 psig 2 valves at 1250 psig 4 valves at 1260 psig urTarget Rock combination safety / relief valve The allowable setpoint error for each valve shall be i1%.
' Oa p
1.2 / 2.2 -1 j
QUAD CITIES
/"'N D P R-29 O
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 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), 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 factor of 1.5 below the yield strenght of 40,100 psi at 5750F.
At the pressure limit of 1375 psig, the general membrane stress will only be 29,400 psi, still safely below the yield 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 the turbine trip or loss of electrical load transients, the turbine trip scram or generator load rejection s
scram together with the turbine bypass system limits pressure to gw) e q
approximately 1100 psig (References 2,3, and 4).
In addition, pressure relief valves have been provided to reduce the probability of the safety valves operating in the event that the turbine bypass should fail.
l.2/2.2-2
-=
QUAD CITIES O
DPR-29 U
i 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-d peak' vessel pre. u e of 1375 psig.
1 Reactor pressure is continuously monitored in the control room during operation on a 1500 psi full-scale pressure recorder.
i J
References
- 1. " Generic Reload Fuel Application," NEDE-240ll-P-A*
j
- 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 NED0-20693, 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.
l j
1.2/2.2-2a s
QUAD-CITIES V)
DPR-29 i
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 be 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 Specifications 4.3.C.1 and 4.3.C.2 and the MCPR operating limit adjusted, if necessary, as reautred by Specification 3.5.K.
D. Control Rod Accumulators D. Control Rod Accumulators At all reactor operating 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 accumulator.
other control rod in the nine-rod square array around this rod has:
- 1. an inoperable accumulator.
- 2. a directional control valve ele-trically disarmed while in a nonfully inserted position, or A)
- 3. a scram insertion greater than max-
/(j/
imum permissible insertion time.
~
If a control rod with an inoperable accumulator is inserted full-in and its directional control valves are electrically disarmed, it shall not be considered to have an inoperable accumulator, and the rod block associated with that inopera-ble accumulator may be bypassed.
E.
Reactisity Anomalies E.
Reactisity Anomalies The reactivity equivalent of the difl~erence 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 conpared to the expected operation shall not exceed IUk. If this limit is configurations at selected operating conditions.
exceeded, the reactor shall be shutdown until These comparisons will be used as base data for the cause has been determined and corrective reactivity monitoring during subsequent power actions have been taken. In accordance with operation throughout the fuel cycle. At specihc Specification 6.6, the NRC shall be notified of power operating conditions, the critical rod this reportable occurrence within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
configuration 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.
Econonde Generation Control System F.
Economic Generation Control Sprem Operation of the unit with the economic gener-The range set into the economic generation ation control system with automatic flow con-control system shall be recorded weekly.
trol shall be permissible only in the range of 654 to 100% of rated core flow, with reactor power abose 20%.
O v
3.3 / 4.3 - 5
t 2
QUAD CITIES DrR-29 C.
Scram Insertion Times The control rod system is analyzed to bring the reactor suberitical at (U
a rate fast enough to prevent fuel. damage, i.e.,
to prevent the MCPR from becoming less than the fuel cladding integrity safety limit.
nnalysis $if'the limitin9 power, transient shows that the negative
~
reactivity rates resulting from the scram with the average response c.
all the drives as given in the above specification, provide the requ. ired protection, and MCPR remains Ereater than the fuel claddin5 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 ODYN transient analysis computer code.
The basis for choosing 0.73 seconds is discussed further 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 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-h able scram insertion times specified in Specification 3 3.C.
d 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 scram tested shutdown.
following a 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 belbre system deterioration beyond the limits of Specification 3.3.C. The program was developed on the basis of the statistical approach outlined below
(
ar.d judgment.
The history ofdrive performance accumulated 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 time is accumulated.The probability of a drive not exceeding the rnean 90% insertion time by 0.75 seconds is greater than 0.999 for a normal distribution. The measurement of the scram performance of the drives surrounding a drive exceeding the expected range of scram performance will detect local variatier.s and also provide assurance that local scram time limits are no. 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 o:her HWrs such as Nine Mile Point and Oyster Creek.
{' '
He occurrence of scram times within the limits, but significantly longer than average, should be viewed as an indication of a systematic prublem with control roJ drives, especially if the number of drives exhibiting such scram times exceeds eight, the allowable number ofinoperable rods.
O 3.3M3-10
QUAD CITIES DPR-29 within the prescribed limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, the reactor shall be brought to the cold shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
Surveillance and corresponding action shall continue until reactor operation is within the prescribed limits.
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) for 1F secs ave 1.44 (P8X8R) 1.42 (8X8/8X8R) for1[bve g.86 secs .385 17
+ 1.109 (P8X8R)
.3851ta" a
+ 1.089 (8X8/8X8R) for.73( Rave (.86 secs where 7
= me n 20% sepam ave insertion time for all surveillance data from specification 4.3.C which 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 g
g Figure 3.5.2 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-state MCPR is not returned to within the pre-scribed limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, the reactor shall be brought to the cold shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
Surveillance and corresponding action shall continue until reactor operation is within the prescribed limits.
3.5/4.5-10
i 9
QUAD CITIES DPR-29 9
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 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 rod-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 than 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.
r i
i 3.5/4.5-13a
t QUAD CITIES DPR-29 ex '
!\\_-
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 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 transitnt, 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 initial value of MCPR assumed to exist prior to the es initiation of the transients. This initial condition, which is
(
)
used in the transient analyses, will preclude violation of the
\\
t fuel cladding 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 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 bypass l
c) Loss of feedwater heater i
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 reload licensing analyses specifies the limiting transients for a given exposure increment for each fuel type.
The values specified as the I
Limiting Condition of Operation are conservatively chosen to bound the most restrictive over the entire cycle f or each f uel type.
l The need to adjust the MCPR operating limit as a function of scram l
time arises from the statistical approach used in the implementation i
of the ODYN computer code far 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, and model uncertainty). These analyses (which are
(N)
Adjustment Factors which have been applied to plant and cycle described further in Reference 4) produced generic Statistical specific ODYN results to yield operating limits which provide a 951
's-probability with 95% confidence that the limiting pressurization event will not cause MCPR to f all below the fuel cladding integrity l
safety limit.
3.5/4.5-14 l
QUAD-CITIES DPR-29 j
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 0DYN 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:
Y = 4 + 1.65 (N /f Nj) U2 7 B
1 ;-i where Af = mean value for statistical scram time i
distribution to 20% inserted CF = standard deviation of above distribution N 1 = number of rods tested at B0C (all g
operable rods)
AN i = total number of operable rods tested in I'8 the current cycle Th e v a l ue f or 7'8 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 formulqting and implementing the LCO, a conservative value for $N i of 708 (i.e. 4x177) was used.
This represents one full core data set at 80C 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 #andCF 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 h
result Specification 3.5.k initiates the linear MCPR penalty at
[/
\\~
a slightly l o we r 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
s QUAD CITIES DPR-29 For core flow rates less than rated, the steady state MCPR is This increased by the formula given in the specification.
ensures that the MCPR will be maintained greater than that specified in Specification 1.1.A even in the event that the speed controller causes the scoop tube motor-generator set positoner for the fluid coupler to move to the maximum speed position.
References
" Loss-of-Coolant Analysis Report for Dresden Units 2, 3, and 1.
Quad Cities Units 1, 2 Nuclear Power Stations," NE00-24146A*,
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 Engineered Safeguards," April, 1969.
" Qualification of the One-Dimensional Core Transient Model for 4.
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.
O Buchholz (GE) to P. S. Check (NRC).
R. H. Buchholz (GE) to P. S. Check (NRC) dated January 5.
- Letter, 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 performed.
3.5/4.5-15 0
Quad Cities g
DPR-29 h.. ::.:=...q. :.:.
n_.:
. 2.:.n....:::t:=...:n..r..=.p. n._n..u
- n....
- . :..... u.n..y..
.n..=... -::::r. ::.. :a.=....
=...... _.=.,.n_..=......,
n :
.......a....
.. 7 nntnn unen:.:::::n::=-:= ::--
.~.3 nn _ p:=.~:.
s,,,g unt::= ::p. = =::= ::n::...I=..i:Tu== ":
tiir=...::p"n::
6du:
W".. r "..
2"."+..=.4.._
12.0..:."..
- =. =.. ~. i:.. ".
. t~1". =.
..=.. =.. * =. _=.a nn-
.......".r4---
- n..z.a..n.
- n..n...;:.:.r.rn... ::n..r..:::!.n.:*.L:n..:...:my::.
g:::+n.n_..n_..
- a.. :: n.r u.,n,.Lnu...:.:.:nn =. =..n..n...
..: nnu.-
4..-
7
.m..:..n..::. :n.. q..u...n..e t.u. n =... :n..
.n..n.2:.n. n.u..t.n..n.:.:.._.t..
.. ~ _......... _
- "2_.=_u n._r4,~._....
. :n..=. n::4m:......... _.
unin:::
- =... i.n....
.. a.. n.:.:....
n
...g.....
....7.-...
..d..:=..: =. a. n...:
.._:::i:5.g:.n..!,n:.::;=:.:.:..:. n..,=.:....:,...... _.
,.....,......=.. :n..n.......
=...:.~:.m.=....~::n.:.
:.
m n..... a. ". g,...
- a..:. n..:...
n..l.n..:.:.:n..::.:.:
_nu.s::
..=
.....4
..g.......
. m.. 2...
..n..n... u......n.... _..
=. ni'-T=..:., _.1..._
...p.
- nn. :n r. : nn2 =g..... _nur= nn nr.n::l===t=:
.p
. ~..
.. -. ~...... -.
- =-
- , n-
... -::::r:= :n:p:n:=
=:n:.;= nn =::nu 11.5.:n:..
.nr[.n:.n n = l...4....i
.l...,
....._...._I*....
;nn t' n=7 =-- +o---- +----=:
=lm.un;= =::.., n t =. _.e.=nnnm:.n:=- - '....nn' =* - - -
- l.. !.._:;n:nx
.... ~..
- - + - - -
..a....
- u h................. : ::nn
. n::
........1..."....
r L... t
- :- = :l=."......... 4
- r 4
.u.... j _..
...:._.,C...,_.. _d.....4..
--::... t
........... _.... nnr m :.,.._. _
4
......-. n.3.
a..
N : g....t:
r.n=. 1........... _....
....g
..g =.
. = un::::
.. m::l.....::=;nn :=: = natun....
- g:::nn:
- n:n~_ nn, r:
- =: :-
.........................u....
.....i.a.......
4...
- nn-m.v.
a.
_:..=n; = n.nd...._..h_-.. =...
.....=...
_.. -_a n== : nnn.---
inn: = n 4
unn e n=:
r : nI...._"....... n. t =
- n==.
.. : r........n.. _n tn =. _=
..............4..........
=:=j ="::. -l:nr..=.
......_.a n:n n.
qnn"-4= __
un=nt== = - - - -
n:- :
'" "= p~:. r inn:: nt= nn
- ".in :q:~r
. :t=. :=
n -
- 11. 0.-
=..r4=nn=_.p.".n=njnn.n=:=nnn-
=nn!==r:=.- In==n
.r t=:: :: =L.n un=j n.:=2.nn:= ::::n.nt.:.=:
i
=1= :
_.n n:=.:=:: nt=:nn: nn:=
=n.
'i' Tir;} 5 ] E % =.d :gn~nanrnn =:=: =:=rnh:d~
- n u:g =:=in n:=. :.-~_ = n
. ::ji~-
=:nnF=;::nm n:n*nnta } =.._7 q. n = r
- mantri -n.::.=1n. t=--t ~~ -
.n ~ ' -
n n=:
rh. " = n..
=:: - =:n u-
- r :: -=,. 2 h c. : =:
-+m. :=:nttm.n.
=
=
=
..n.. 3_.
n.. n..n..
.. n_ =._.....n.=.... {: n. n. =..
.=...i=. :'n_t.:.n. nt.=..._.:.n.n
- n. :. _.=_. 1.n_.u.=._n.=.__=., - 2.n.,.. n_:..__3.._,,,,,
a_. = = =
tn
.....u.n.9:..
...=g..=.,._..:
_=.._.:.gg.......-.
g.
"".. n n.. :.y=.. :n..: : =.
.- n..l =....
... -. u.n..u. n..
=.1:n..t.u-nn..:C.....=..n ;._n.n...
t
._..L...t_r.,. n _=,. r. n. - =_..
n
_: -..p:. =... n r..
p g.
......g g...p....
..,o__
y O
J U.,)
_.3.....
...._.._r,._._.,._...
e
_. n_ _. a. - t = n=
- r==r.=.=..
..;.... =
= n= : =r= -. __
v 0
=. _...p". __r+ =_=._1=.=n..
=..n.. =.. : =._._=ne..
- a.,.
-m.-.=.:=_=..=
Quad Cities Unit 1 n
1_
_r.......
o.
g_
,_m_.. ~ " -t..-.: _ _
._...._+-._-2.._-_... -. _ - _ _ _ -
_. _*.r... _. _ _
.....-.....r._-._.--.
Fuel Type:
Barrier LTA go
,3 e, %
.y...._r_.._..
~ - - -
g nn+=p_n..nn_.nt =-- n. :n:.-*n.; --- -- - n=:-
-n =.
o,
.C %
== = e nu :qn =.r
- =l~.n 0.
-...,... =... =. : n..=_:tn"
.n:. :.
U-a w
1
......u......_
-3...._.
._. ::.. a _.
..._.a..
2,
...p....
... _...u..
a w.
g O,
....n. P. !....... } M...--h..,.... J.~. in.=. :. t =.=. n..j n..i_-h. n. :.j M'.*~.. _n. :.n i=_ _:.=... =..".;ij r.i.i.j - p.i. _5_5.f.?f_in. n nig_-* -_". 3, d 0 4
gp
~n.=nt=.:n E rn:_.nbn:. _n-t.rn ;
..:r.; n:. nn=:nn:._
b; C CO
.n.. : n_. : n.:an nn - -. :nnuntunnut;-. =du:n =..! n:* =
n =:I= nr 2:.__.=n:n:- :==.n r n=+ = --L n==::un. :t : :-.-
r-i A
. 1.. T. c a. :=_r 1 =: :
.=:nnb::r:=1_.___.L.._._.____.4.._....4.-.._.,__-_..._n_._...__.4..__.....__.__
." :.". l...
o 0
f.,
.m.._
rr....._g...
.,.. _........_31,.._.... _....
..._}...~__._..._........_._:_ _._... _....... _.
......_._._..4._._g..._
C 9,5 p
..... _..._p
, o p ;j 0
~
10,000 20,000 30,000 h6,000
{; &
Planar Average Exposure (WD/T)
..........._..L'....._.;_...._.+....._a.._._.._....._.>t.._.._.._..._.._..._..._.a
> c-
........._L..
.._..l......
... _..,.......1_.._........
- ~~-=g._._...,.._1._..~^"----'~-'7...._.
---...."1 :.,... n:
-u - '---" -
et C
- n.=L._.t._u:nn:
- gn.
- _
%gn=i.. ;=1 =nn.t
- = t.
- : nT. u= W:.; j u.-,=:::m...rn.;.n-.
- 12. 0c * - U= = W p:.=:n:: =n:
- =---t-- =l_==7 'A o
. r nL:
r:n "..
=
- - n = =-- - = = a -
- = = = :- P =+2 --N "r : r - =-
po 4
=
n n m.p---. = rn=. n pu.-
- L n =_ :_. -- = h :. =
e ;....= :n-n= :.. ---R=r'%: j.; =::; --: =r_:.n_.: nnn n_.
.i. ;r. n: n=_:t= :n-h 3.-- =: r -- -- /; 2 r n:
-f =:: =-
'r-r=_:=.:..=_.=.a:nnf =: n-a rn r.m.=-
W i.=: =2'-. = m:;L 4_....
__... i.- :n.
2
.n -. __ ::..:..n.. i. =_.......
p n--
f~= nr.....: r........ 3
..=._n_=..=., -.. 22 : ::. =._r f rn.... b. =.. =..
- _._ J.__... __ L.._u.. r.._I,n.. =_=. :c__=_ ;. u._.
-n....
n.
a
.2..
y
~
~.., c..F. L..3. __
_: =.. =. _.U. _m_._-- U. _ _:._.._.J. _.. n.,.r n. -..._.._.. _
._n. n..t =_ = "..n.r.n. r_. n.._r_!.
.. _ _nu.._.= _=.. n:. _: n ; :.: =. _.
i-P.:
i~..
L
... -... -........... ~....... _ _
...._1_.
..._j.._._._.,___-
4
._.._.._..:1......__.J..._...
_.. - n;1:-- : - --
L..,__......-....a..._...
- ur - 7 r-- :.= - --- :r {. _. _
t d...,......;*--T_._---
_-+:-
........___..... __...-...=.._.. _...m.. _m..._._.s_
~. :::. n I
gy -.
..-.n.,
- 11. ye._.
.,.._..4-... _...
."._'.:----*-.~._-~l._.
....4..C--~'_~~.#.._._4--...
... -. _. _.- -. ~.
.. -.... _ _..L....._.
....._....._..L._...._._._._.=.--=.L......_._-..,.-..M.....-.
2
-.._;3__~..4.....
_.*2--"'.-2==
-*--- ^---
s DE
- u--
.. _. ___..._.e.
._...'C.._.
............ -.. ~.. _ _. -.
.. ~. _....
.. ~.
w
..._.t.._J.._._..... _ - _
gF
._.a_\\1.. - __. -.
.._.1..._._._,....._
_._. {=:1L_ o _g._.i-i._.3,
. _.: = r = 12 = =_a. = I m W
w
,._..._.._....t........_.___!_~..__.3...-.
.- 2 :=. 5
.._4...tr n.-_n4-T t= [-
t-7 g=..::;r:l... _.
3
- +nu?= =:
c.W.
_.._..L..._=._..__...,_.=_.~.._._..._.=..._..
- ..I_.._.a......__.._u_......_....--t..-.~.
11, O.. __. =._. _..
- r..t _.___...
_..__t...._._._.-._..:...._._4..~
>J
_._.t,.__..u.._...._.;__.__. _ - _ _. _... - - _ _ _ _ ~ -
4...._... _....
a w_
...._....... _ _..., ~ _
.g.__-.....
_2
=. L=. = 4. =-.3.. NJ =p.......
.,..l...,.=...tr.~_t.=_,
______._m.
4_
o L
o l =_.n.1".
=..=. a
=.. =,. ~..._ _= r_ n.=. :... j... : = c. a..n=--
2 nr=. =.. _ =-
--,_u-.
v g.
- - - -. -: r-=
nnuninri~r
=f "-- r 2
_-t =1==n ; =:=
=_=r.32 -1= ;=j==n....,------........-...m..."..t.....-.....-
..:. =:.-i: _: rCr-- - 3.u n"n=-t.==
m==f=
--m==
nr$=:rr.n. 7 t--
. n== : = =21rn.
r =rr D
r : r: r; et
..=__...__=..=n=.
._~..a. _.. n. -}...- 4 g
.n.-
.n
..t r :-- -:n. --=g-
- -. =..r -..=: _.=.__ u. "_=1 - T.... *.--+.=..- : === ' n -- =
-... g un e n.- -.: = =--
x r. :: _=. t u e =in==_}y...-..._1
_ _. _......a.r:-*... - t ~=- : =' -. -
-,-m---'
- : ~~ - -
...a..=_.=_.L..
6
........_..?.....
10 5 1_... ;.
. cr a,-
=
2
.:t:
r.--e = = --.= -_
n lETT INFE-f f-Ei "_#55iNQf=F:"EEifnE gri--. b. '..
t.....
Quad Cities U: lit 1 ir r-+h_...-...._.-..
..... 3F= F g.
1.
d=.Eir1E.' 12'
- .=J
- . u:
r t.,..
- n t=u =u;=.= = =. u m. =F==u_
- ut- -. ~ =. === :., :
.=*
r--+ =-:==m
...."n_...
I u.: =:. :
f.=== u Fuel. e:
8DRB265L and l'_.=.._..=_4..._
==4=i-
= =: = m - r===:
.=.
10*0- "= r '
+3 P8LRB265L(dashed line) =i a= =
2 =n.==F=
.=.=g=_=:t; ; z ' %n :
.==.. =_. 4 =.n... :r....
... -. _.....1
---r - = :2.n = p= nr :......._ a.r 2.. _..... _. =a =. =.=-..
.... _....... _... =
u.2.....___...,...._.
. n=_.p=. =. 1 =..=.. r
~
===:
...s.
- -: z, T,
---o37...._.*-*g.-
-m.e.....
._..._......s.... -. -, -....... -
-..~.--n......
--3._-!.....m~--*-----
~.....,............_..1_..._..
._...~.1..,.....
..I.-..._-.--.-.-.-..-....-
o _.......
.......$_.-.w..-.....
=
_ ~.. ;
=========p==:==-==.==_--2-..=.
~ - - +..
.-----.--.=n
......._....._.......t.....
g.-
. ~..
.........-...g......_.... _....... = rra=== t r --- =
.....u 95
=:2=........
20,000 LO,000 Planar Average Exposure (MWD /ST)30,000 0
10,000 Fig. 3 5-1 l
(Sheet 1 of L)
Maximum Average Planar Linear Heat Generation Rate (MAPLEGR)
... _ _ _ _ _ _ _ _.. _..._ __.vs..Ilanar Average. Exposure _. -
1
.s p pg.rm.
19R-20 23
.n...::.
...;;.: _...u..
.w.. -...r -e.
. -.- w.. -.,w.. u
_L _ _
+: a -.-
..::.[_. J. J. _,.
- l. _ ;.1__ _. _.I,....(.,,i
..f....,. _4_
~
.l l
8..
- m -..
-e a
- i
.... p... :. _.... :,:.l..: ;...;..c..n.,.. ;.
- ..t..:.g.__c..._
,,'b' 6,
i t u !.-
} g,
, ;_ _2
..t
_..: !.. ;... 1 l.; 3 ; _ _. '. "..
. '. _.I '...:;..-
4 e-g g
a..J' g
.g
.:. ~
._ q,.
.,. g t..
....g.
_7_. -
i
...a.
4.,.
- ...j s-a :.. e._ _.: :-.: ::. -
3
_ :..=._:
.r.:,;.... _. - _,,
.5 w..... :._.. - a.-
... S.s _j ;._ _ _
3
__; _?_y. _-l. i- =, c.~:--
- a
.....:..r....
- H---
. l:-#..,7
.,_ - r - ':- - - m - -
4
..'.a.-./.l.i.".....
_-.._t.
-... #.. L
- r
_t r : -: a:,r,._.. _; r.. w ;. -
,r m.-
- as,
...l..
s.
se.. 5 b.,
g.'
10-
......_e_... _-.....,..,.,.
=..._.-_=__t==_a._:_.
Quad 0ities Unit 1 m _
-cD o
.u 4. r.r..
ny-FLr,
e r-n,
- u. _
..._._.__._-__._-:-r,.--==.__-_--===-_..-=._:.=.==._.-==--=.=.-_:-==.:==_.-__
=
=;--
s g.
T :=. :::: m.: : = =T
- =
- =- -
ef/: 5 - {
r--r- =: i- ;.==._ =.
3 0,
0 10000 20000 30000 40000 2
Plara Average T.xposu e (KC/S"')
8 13 a.
-:-:...._.__=___..__--- a... < _.. _... -
.=
o
....s u
.. _.... L _ : _.,.)
....-..~._;=..m..,a c.
.a_.._.___:.
_...-._r..-
__.=._..._..._=.__...a.___
~....:._._...,
- _ =- 1-1 _".Zri-A-'
_, ~-.
.M.L, *g.:!g.. _J. -J,.;-d2,"- -.....' -; _,,.2.=-
g
=..
g 0
12
..... ' ~ '.
~."
. l "-
et..
e.
L.
t u
.. _.::.. g
.a & =_ _. r....m..r.. _-:--.=.-. -.
C
. -.- - s1 2
?Jp TIF4, E:E ;:i.:.- -. - - -._. = i -L '-.; -"M-l.,--. 4 l-; 7
. =,'.--,. '
V 2
l
.1
'_t s......._
.: a.. -
.....:, y _. -:
'..m.,!.........
ue 11
. i...-
c
..-..... _. _..~
i...
e
..4
.,. s
.-li i :.s. : _ e g
E
.'._.1_.-
7_.,
- - -.. g ;._., _. ', '.
.. ~..
6-g
. m -- _-
e
- J : : !.: = _. 9
- s...
..._ :: :4_ -2,- : ?
-j;,.- - _
- 2
?-
--:"3
---2.
e w
.a. a...
.I _ !
n
.,,....s.....
w l
.iC_.?..,. :._...... _..
sr>
Quad Citics Unit 1
@ =.A- =.:.. - :t
- ..r.
ggy wg. py5g
- .=._...........
g....
x.
.._ w.
.: c.....
.:._,_..........y_..c....:
... : r."
- .... : ;....a_..:
l 9
0 10000 20000
/ 30000 40000 Planar Aversge Ex;.osure (K.0/ST)
"inute 3.$-1 h hV" ^V"I'w Planst i.incar "e.at f.cneratbn
(..tret 2 of h) trat e (MI.'C") W.. t'l aint.uimo 'pur e r
g Amendment No. 61 Y
M 9
WM
. =
..p w.
Quad Cities i
DPR-29 ti:t+.r*:.:r'ta!..+- ~ - - * * + - * ' =t it=:+*~: :2*-- ' * -.
- -~"i
=
Li. t'.t.:t.*:.. 2:f.@
- g y;;,
,4,,g t+-
--v 4---
- grt-***'*~-~i
- r ".^-
~.4
,~~*-+
--+-+
~ + -
9
.p g::::n:
tn:.._ n. ;.
4---
- =:-_.
.3="*.
.J. :
.J 4
. - * *2 g
....-. +.
-"n-* =n::=:tu0 =*n =
...2=....._.
~.I f 2
7*:::::::*"-*-*
+;41 -
- i =:* 1="t*- * ^
r.
---. ur.".:: :;
=:.;r~~~.-*..N..=t.~
"-- --, ~ + - - - - + --
- n::::nin::;.:;-
Y
- .=$*= *::' *-
^ ~; +
e a
_4
.a..._.4
- ==:t= = ~- :-r-t
--" - H - -.g =.:. :.4 =::
a a g.
._L 1: *~ ** * **d -*** j** t I'**-'"~
~
-- - :n I
... _ ~
- _.2.
1
= H ~4 *:-/.;rn:.._
L :... =.". :: :==t=..
1
+.--
- - *, - ~
l
- 23. =*y'
- q....g...
..~d..,..^5~
i
= - g- =L =;.,=:=:== _.._ 4.
._*1=:==.
x i _,
s
==
=*
-e-
- + --
w=2==
-.a_-
s i
,= : = =: - -- = = =:== = = r -
=- ---: =d.--- v-
~~
- =:.=::=:=t =f==:==:= --" 1
_ -t n=:== ::-:.= =
. = =-: = r =.= _
~ ::1= nrE! :* =l...._
.._. ~._..-,
a..,.
-,.EE EiE~..L=; =
t
_E: =2===i:
h--t== --- -
=2_
,..y.__._.
---r:--~--==..+___= ur.=fy-
.:_._.:_..= = =.==-.=::--" "._
._ _===. !
r-
= -. -. - =. ::=-
= :._, -= =t...
--- : 2:-i
._. -t====t-
=
- w. _2==-+-=~___
_3 1
4.==._L.
_=n. _ m _==
_ a:-r _ _ :r_ __;-. =_
17
- = = r.= = = : = = 3
- ====--- __=-
-r.=-,.
---._.=t.
1 a==t:--t
,-==:_._.
r_.
- = ; = e : := = = E====:===n.- =-- -- _._..__.- - - -
- --:_.N, s=..-
- --+
,- r
- === r
---4===
r b
= = = = = = ~
w_. __
o
==-.r=._,_..._
m
._.._g--
= = = = = = = = = -.
Qued Citles 1 i L____ _ -. _ _....-
s.
o5 g,
===2 :==-r a__...
a e r- -
- ==$... -a-- -
c
-2:=--
=: s ns
. o,
, o Fuel Type: P8DRE239
._- -. w -
' =e4 y.
4.. _ g
.-L 1 - --*
e
=._=.:*--*=3
~*
- =
~
2e e
- =:=-==__
u; a
=_....a-_-.._.
~ =--+
e 0-C
=..+_.-._2r=----,_-,-.__E_2.
+. _ _. _.
.a_. - _- _-_ _.._._.=
c
. =
y e
= = = _ _ -. _.. _. _. _.,
o
-=
__I
+'
==--.==-!.____3=_=.
- - -- - - ~ = - - - --
-~
e 10
- E 1,000 20,000 30,000 ho,000 Flanar Average Exposure (6'D/ST) g(2
- z.===
. _.:. _. __. =r v
n=.=- ::==- -r - _4===== =.. _...=.a
_:___..==-2=--==
es u---+-- r -
c ::
- = 2 :
=-
.. _ _. _ _.....__..__._._._...---_.i--
t==-..._.._..=_=.=.._
-__=.-.=...._.:__K...._.-----~=. - - _., _ -. - - -
.._. = :=...= =.;
. = _ _ _.._ _.
< 's.:
~
a, g g, g. =. =.=. = = =. =_..._.._._yt..__ _
-- = :.
21__---...==-=+.=__.x.+=
.==---,:=
--:=___-_
ue
- .__.====:-
eo
- = = =- =-r e---
_. _.. _w..
c:
- ._. :..=._..;v= _
=1;_ _,. _.. 22 =_.._-
_a
. ===._ _ _ __ _ u -._.__-
=
eA t =--==
= ;
--;.==_. - - - - =. _ _ _ _
S
[5
- r7 _
- =- _g==.=_,=_ _=_ __.. _. _t==.
=- - -, =...
==
. _ =
g
_=:- =__
- -._- __ 7-==rgg=
_ g ag;g, _:_
oO n;ggp7===_m.: =_
11.,1._._,_-..__-_;_.__v._-._.,_..-_..--.
g -- -
_=;.;_-.__,-_.=.._----.. = _ - -
n, k
- ==.====,= r =;== -:=_ =-... _ _.
2 --,2 __.
- :=:_.
~ - - -. =
- - + - -
a k
u l
c.e e,,
t
~
e-
=_ _ t _-__
4
_,-4
.2--
4
+ - * =
l yy g
=-
9" 3 g
1 a._
l e.J E
6 a2 11.0[_ _ _ _ _ _ _.
4
.- - * =, =. _,
=-==-
i
-1 x
e 3
2:
.=
_..___a._
o
- = ---
_ _ _ = -
+-4---
N =.-
- __.__ E1 _ ___
E r_.___,i...__.- --..__.___.A=-._.._
w.
-.. _..-- n-w N
10 5 ;== =.=_ =_____. __ __I= -
- - =
- = = =:.
x-t.=_i__ ;= = r
.-_=r-l g_;
.-N=~-
.===
=== r== 2 2 -
Quad Cities 1
==-t==N,=---==
-.___.__._,%---*=_-.=:====_.
.. _ _ _... _:=========.-- - - - -
Fuel 'Npe: P8DRB265H a
10.0;___ __-'~U*~~~~~~E
~
"^-"
l
.:=---__..__ _.
_ _ __ s. -
_;. _, :-.- =_.
_:= _._l..
.. = - - - - -
- - = _ _
-___-+v.-
_ _ _ __ = - - =.
l
=:==r.A l
__==2=
- .- f
--- c=4
_;.. = = - - - - -._.
..r=-
, --- L0,000 p.5 0
10',000 20,000 30,000 Planar Average Exposure (&*D/ST)
Maxitu: Average Planar Linear Heat Generation Rate (MAPLHGR) l vs. Planar Average Exposure (Sheet 3 of h) Fig. 3 5-1 l
$a ll'. #.
.=.a.a.-._.m.....m o-7
i DPR-29
..e,. ]d M :! :1..
! ntti:n.j=.
.... :n4 = :,;
.MO.
- Mn : :n"l::
..: j:=!!
...::? n1=a n:......
- W Cin....
"dI *:
- 1::.:
- !". n:: t-* **.t :nn rn
. l::*!
IMO. Ul t * -
n!=
":y.". b.....
a :
.. j **" t*7*:.
.~.J..
..j"......7****. ' '. ' ' ~ * "..
'.#."..u i:-*; ****
4.
._*i. ". * "....H.". 2. * "..l. 4...
""I..Y.
2.
"..#..*.I".".'"
".',.".._'I."..""
}2,Q
.".. "...;;l". '
' ". F. *
... ". ' ~ " ~ "
i h:4. XII{;*!"j".-lie.ii!Ei M.pE!!!n**** fiE
- iE}EiE515:EM hh55 E!NditiSihEilEic Elii
-*=::*:
-". f": 1.. n...
.t d
!t _.
.3.
.*"*n
.n1' M. "b
- 3. :';. =...' ;."f n.:l..,
_.............,!........~.n*,".n.
- =. l *.::.. * *.a. : U
........ in'.n ;I... 4
(' n......
'u
- n.. *~* } U..n
- I.
- . -
- -*1.~.*u..
... n.n.. t. :.:. ;.
l
~...
.. )
.. g
.. ~.......... ~. _..
... -...g*.....
T:*M '
_..f m. ~
. :.....!T 2n.*g'n:!):*nU -
- d * *
- nj:'="
.:::L;N: 2;;. ;_g....j.." " j ::...'....
..a....
- ':n..*. :~=., *n..
- 2.... ::..r. ;..
e
~r
- - l.~
o
--r-n "
+ - - ~ ~ ~ '
-~.
p
.4..--. uj;;..=%o:=t = *l=.:.m... -
- =3=u nun
=:n 1:n!:=1.n4*= ;n
- n:
.r
- .n !.:.. " "M 4 r n'"" :n-
~' m m
11.5 d
"= " : '
u r: " " = ' = " " - " =t n
~'==a=
=1
t=an-NiiNh'$iNNd ' ii;N5b k E2ENN5NfEd i!UN hihNf! b b!!E Nn !5h NNbNEh55Nh!5:5 C
=
3 ii W P inili B-!iH= t=@illiidiMiMEE!!MUs ~:iiM;. E6MHt=l%!#EEEi# Eis= 55:E_iEi=iisEHEJ gn:: :a;= :.:nt = :..nt.= ;_. ' ;; :n;;=:= :=M:n+::n.:nni:--" ::
.d
/n.q.;
.:4 nan.. n.:. n. :..... =. jn. y =: "::"=M.= [= : --- r-t r tur; ":.:nn"........ " _
7t;=l=.. f ~ nt u : n : r t. :..:n. :
- - t - -- :.
4
.=. ; " :..
c
-ch =1::..:= Inn.
n ". L....t.....n..
nt-
....E
..o.......
e
=:=t::a..a.: n:=.j,n ===n=j=~ I=t=t =-*- --.... _,n= : -* - ~1_? " =; t = =...
e 11.0
- ~ ~ "-"--"3=unn==nn j.:nm.".;n.N:= - - "..":_ n r--
'-".a~~~
.un" rp=. nt-...:.Lg.";p-:n n"n:r--
.u n =_==. ntnnn=
7*
c.: = L..:*:" j:.:.
u j=...::
e
. - ~ " " *
- ~ ' - ". "- " 7:3 =.An =:=t
-"2n=
==
nTr
.. - M T.n ' n, s._.....
- n. _ i=n.V n= urn =2= ta = :--1
- = p.
t.= p~c: = =n; y:*:-n- :,".:32:jnc:=n::ua 4
=:=..
M*9 -" - n.=~ : : : n"-rt:r. t =r=g_n r t-- =r t---" g nr.---t-x:2:n n;=1 d
=p
. r== t.=r=r.U : t =r = =: nrm== != * =f. * = 2....._..j = :=% =nnt:2
=1..- ta I
,..;: j:. :~7={ c=n.=:....in n.:=r::=i=:=
At.
- . - -* {
. = n. r-n
= * = t i 7'" : = nm...._.
=.. * *
.r-t................_..,
._...~....-..g.......__.g.............._.4_..
_.._.._.....1 b
- - *- -- - - - -- - - - - - -* - -.. =:~.. n n..n. 2. _n. e.n_.r. a.t=._ _n. _.t..~1...~. ~... 'n :1 =2 =....._r. _: n-..c.. 7.n*=_. a.=.. ;,. _ 2.
1
. n.M *- :.=1pn,=: :M=tn=.
~
7 d
=i=n- :_~.. y_ :===nwr==n== = j o
e-10 5 y.. _.= = = = = i
=
r :
q =:
t=:
C4 f.=.. : t n - n n"...!"" -
e.-
a s.
=
q..=.-n==..
Quad Cities 1
~l.=.._......=....=_
.... N_ =.. *. _:.
....T.......
g.=.
O._= u.=__==..-___.i
=..p.=.._=_.
4 oc 4 44 oEe 4 *
- =.
- r=n;;.x = : r=;=
- .;; nn=_ t
-~ =:m:n.*
=--- -\\:==_n=
,3 e
=.u----
- =2 =.
- n. ;
_n e,
E*
S._ =*. =:=4 ".n c.
Fuel Type: P8DRB282
- r*==n--
,_._.:.T==N== us
~
=
"~"
"#~='"
"~
u-
."...=.. _-[.:..c n2= ia._
._r_.=_E=.._=._
10*0 ri, d_=._3 _r=nt._=A....=s_
mr
=.. -
...p-w
..:.. = r = v w._.,. _...__... _ _ _.._ _._..._. =_:4_____m _r... _:r_._...n;=_= = e nzrur.._ __,._..
....._..x, t
e.
e
.-H..2 3_.... _ n
,=...:n._=_t=_.=_1_=_.._E..=.._=_...____..~_.._..
g c
....__.._-.an_.u_:=....._=.T......=..__m.-==:
n._=__=
.. _.... _ _ _...... _.. =_ =__
.=_=2._2,=.._=.=~..=.
y n
.m l==:c c.1.u =: r.= j =.:.:= x::.--...= n--.=- :.--d :== f=--" =2n==n : :.:. -. 4== =).=rrj======r--
--" " ~--
- n p
===t==-
=r===-
~===2===-
=
===u="=2='~'" m r-9.5 10,000 20,000 30,'000 Lo,000 D
g dr Planar Average Expecure (!GD/T)
_5 Maxi::.um Average Planar Linear Heat Generation Rate (MAPLHGR) vs. Planar Average Exposure (Eheet 4 of h) Fig. 3 5-1 l
1 D
w 4
L E4 It C it w
2" w-a c2
- a. -
>w._J I
4 O
t 2
I
=,
f5 4
>l 0
Z t