ML20246N655
| ML20246N655 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 07/11/1989 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML20246N654 | List: |
| References | |
| NUDOCS 8907200015 | |
| Download: ML20246N655 (43) | |
Text
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ATTACHMENT 2 fEQEOSED_TECIINICAL SPECIFICATION CHANGES A.
QUAD CITIES STATION UNIT 1 (DPR-29)
B.
QUAb CITIES STATION UNIT 2 (DPR-30) 01887:11 8907200015 890711 PDR ADOCK 05000254 P
FDC
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QUAD CITIES DPR-29 I
l TECHNICAL SPECIFICATIONS APPENDIX A LIST OF FIGURES Number Title 2.1-1 APRM Flow F.ference Scram and APRM Rod Block Settings 2.1-2 Deinted 2.1-3 APRM Flow Bias Scram Relationship to Normal Operating Conditions 4.1-1 Graphical Aid in the Selection o' an Adequate Interval Between Tests 4.2-1 Test Interval vs. System Unavailability 3.4-1 Deleted-3.4-2 Sodium Pentaborate Solution Temperature Requirements 3.5-1 "exia.a Average Piener Lineer "eet Ceneretien Rete '"APL"CR) bele+ed v:. P!:::: ^;;r:;; E:;;;;r:
3.5-2 X feeter beleted Mfnimum Temperature Requirements per Appendix G of 10 CFR 50 3.6-1 3.6-2 Minimum Reactor Pressurization Temperature 4.6-1 Chloride Stress Corrosion Test Results at 500*F 4.8-1 Locations of. Fixed Environmental Radiological Monitoring Stress 6.1-1 Deleted 6.1-2 Deleted-6.1-3 Minimum Shift Manning Chart vii Amendment No. 117
,1, 0 DEFINITIONS The succeeding frequently used terms are explicitly defined so that a uniform interpretation of the specifications may be achieved.
A.
(n tW)
B.
Alteration of the Reactor Core - The.act of moving any. component in the region above the core support plate, below the upper grid, and within the shroud.
Normal control rod movement with the control rod drive hydraulic system is not defined as a core alteration.
Normal movement of incore instrumentation or movement of the TIP system is not defined as a core alteration.
C.
Hot Standby - Hot standby means operation with the reactor critical, syst3m pressure less than 1060 psig, the main steam isolation valves I.
closed, and thermal power not exceeding 15%.
D.
Immediate - Immediate means that the required action will be initiated as soon as practicable considering the safe operation of the unit and the importance of the required action.
E.
Instrument Calibration - An instrument calibration means the adjustment of an instrument signal output so that it corresponds, within acceptable range and accuracy, to a known value (values) of the parameter which the instrument monitors.
Calibration shall encompass the entire instrumeit, including actuation, alarm, or trip.
Response time is not pa t of the routine instrument calibration but will be checkeJ once per operating cycle.
F.
Instrument Functional Test - An instrument functional test means the injection of a simulated signal into the instrument primary sensor to verify the proper instrument response alarm and/or initiating action.
G.
Instrument Check - An instrument check is qualitative determination of acceptable operability by observation of instrument behavior during operation.
This determination shall include, where possible, comparison of the instrument with other independent instruments measuring the same variable.
CORE OPE RATidG LtHITS REPORT -
The unit specific documenf core operafin3 timi+s for %e current opera +in3 r
hat provides reload cycle.
Thes e, cycle specific operafin3 limits shall cort be determined for each reload cycle in accordance wi+h Spec.ificaf ten
- 4. 6.
Plan + epera+ ion wiWm these operafin3 timi4s is addressed in individual specifica4 ions.
1.0-1 Amendment No. 114
f c
QUAD CITIES
,)
OPR-29
]
The MCPR fuel-cladding integrity safety limit has sufficient conservatism to assure that in the Event of an abnorma1' operational transient initiated from the normal operation condition, more.than I..
99.9% of the fuel' rods-in the core are expected to avoid boiling transition.
The margin between MCPR of 1.0 (onset of transition boiling) and the safety limit, f s derived from a detailed statistical-j analysis considering all of the uncertainties'in monitoring the core i
operating state, including uncertainty in the boiling transition correlation (see e.g., Reference 1).
Because.the boiling transition correlation is based on a large quantity of full-scale data, there is a very high confidence that operation of a fuel assembly at the condition
-of MCPR.= the fuel cladding integrity safety limit would not produce boiling transition.
i However, if boiling transition were to occur, cladding perforation would not be expected.
Cladding temperature would increase to approximately 1100'F, which is below the perforation temperature of the cladding material.
This had been verified by tests in the' General Electric Test Reactor (GETR), where similar fuel operated above the critical heat flux for a significant period of time (30 minutes) without cladding perforation.
If reactor pressure should ever exceed 1400 psia during normal power operation (the limit of applicability of the boiling transition correlation), it would be assumed that the fuel cladding integrity safety limit has been violated.
In addition to the boiling transition limit (MCPR) operation is constra,ined to a maximum LHGR = 17.L/ft for 7 x 7 fe; i.id 13.nt!/ft
.m_
e o.. _, " --
This constraint is established by Specification 3.5.J. to provide adequate safety margin to 1% plastic strain for abnormal operating transients initiated from high power conditions.
Specification 2.1.A.1 provides for equivalent safety margin for transients initiated from lower power conditions by adjusting the APRM flow-biased scram setting by the ratio of FRP/MFLPD.
Specification 3.5.J established the LHGR maximum which cannot be exceeded under steady power operation.
B.
Core Thermal Power Limit (Reactor Pressure < 800 psia)
At pressures below 800 psia, the core elevation pressure drop (0 power, O flow) is greater than 4.56 psi.
At low powers and flows this pressure differential is 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 speciDeJ in 4he CORE OPERATING LlHITS REPORT for veious fuel types.
1.1/2.1-7 Amendment No. 114
QUAD CITIES o
DPR-29 2.1 LIMITING SAFETY SYSTEM SETTING BASES i
The abnormal operational transients applicable to cperation of the units have been analyzed throughout the spectrum of planned or erating conditions in accordance with Regulatory Guide 1.49.
In addition, 2511 MWt is the licensed maximum steady-state power level of the units.
Th"s maximum steady-state power level will never knowingly be exceeded.
Conservatism incorporated into the transient analysis is' documented in References 1 and 2.
Transient an;1yses are initiate,t at the conditions given in these References.
The scram delay time and rate of rod insertion allowed by the analyses are conservatively set equal to the longest delay and 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 tne 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.
'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
..-+ Peregrept. 3.0.K as the limiting condition of operation bound those which are conservatively assumed to exist prior to initiation of the transients.
the CORE OPERATl4G LIMITS REPORT x.
1.1/2.1-10 Amendment No. 114
t
'4, QUAD-CITIES DPR-29 TABLE 3.2-3 INSTRUMENTATION THAT INITIATES ROD BLOCK Mini.num Number of Ooerable or Tripped Instrument Channels per Trip System [1]
Instrument Trip Level Setting 2
APRM upscale (flow bias)[7]
~(0.58WD + 50) FRPW G%
2 APRM upscale (Refuel and 112/125 full scale Startup/ Hot Standby mode) 2 APRM downscale[7]
>3/125 full scale 1
Rod block monitor upscale (flow 1 55"D t2E2]
b3 0
bias)[7]
1 Rodblockmonitordownscale[7]
- 3/125 full scale 3
-IRM downscale[3] [8]
>3/125 full scale 3
IRM upscale [8]
1108/125 full. scale 2[5]
SRM detector not in Startup 12 feet below core centerline position [4]
3 IRM detector not in Startup
>2 feet below core centerline position [8]
2[5] [6]
SRM upscale 110 counts /ree 5
2[5]
SRM downscale [9]
110 counts /sec 2
1 (per bank) High water level in scram 1 25 gallons (per bank) discharge volume (SDV) 1 SDV high water level scram NA trip bypassed
.Y.
3.2/4.2-19 Amendment No. 114
4 '
QUAD-CITIES OPR-29 TABLE 3.2-3 (Con't)
Notes
[1] For the Startup/ Hot Standby and Run positions of the reactor mode selector switch, there shall be two operable or tripped trip systems for each function j
except the SRM rod blocks.
IRM upscale and IRM downscale need not be operaole in the Run position, APRM downscale, APRM upscale (flow biased), and RBM
~
downscale need not be operable in the Startup/ Hot Standby mode.
The RBM j
upscale need not be operable at less than 30% rated thermal power. One channel may be bypassed above 30% rated thermal power provided that a limiting control-rod pattern does not exist.
For systems with more than one channel per trip i
system, if the first column cannot be met for one of the two trip systems, this
)
condition may exist for up to 7 days provided that during that time the operable system is functionally tested immediately and daily thereafter; if this condition lasts longer than 7 days the system shall be tripped.
If the first column cannot be met for both trip systems, the systems shall be tripped.
[2] Wn is the percent of drive flow required to produce a rated core flow of 98 mT111on Ib/hr.
Trip level setting is in percent of rated power (2511 MWt).
[3] IRA downscale may be bypassed when it is on its lowest range.
4
[4] This function is bypassed when the count rate is,> 100 CPS.
[5] One of the four SRM inputs may be bypassed.
[6] This SRM function may be bypassed in the higher IRM ranges (ranges 8, 9, and
[7] Not required to be operable while performing low power physics tests at atmospheric pressure during or after refueling at power levels not to exceed SMWt.
[8] This IRM function occurs when the reactor mode switch is in the Refuel or Startup/ Hot Standby position.
[9] This trip is bypassed when the SRM is fully inserted.
spec.i fic3
[.lo] The Red Bt k Hon *+or upscale se+p. int shall be esMl;.hd as in 4 he. CORE OPERATie4G LtHITS REPcRT.
3.2/4.2-20 Amendment No. 114
< :. L QUAD-CITfES OPR-29 interval not more frequently than 16 weeks nor less frequently than-32 weeks.
These tests shall be performed with a reactor pressure above 800 psig and may be. measured during'a reactor scram.
Whenever all of the control rod drive scram times have been measured, an evaluation shall be made to pro-vide reasonable assurance that proper control rod drive per-formance is being maintained.
The results of measurements per-formed on the control-rod drives shall be submitted in the annual operating report to the NRC.
3.
If Specification 3.3.C.1 cannot 3.
The cycle cumulative mean scram be met, the reactor shall not be time for 20% insertion will be-made supercritical; if op-determined immediately_following erating, the reactor shall be the* testing required in Specifi -
shut down immediately upon de-cations 4.3.C.1 and 4.3.C.2 and termination that average scram the MCPR operating limit ad-time is deficient.
justed, if necessary, as re-quired by Specification 3.5.K.
4.
If Specification 3.3.C.2 cannot be met, the deficient control rod shall be considered inop-etable, fully inserted into the core, and electrically disarmed.
5.
If the-overall average of the 20% insertion scram time data generated to date in the current
& lt a specMied la %e core-4 cycle exceeds 0. n ;eeende ythe oeERATids LtHtTs REPORT MCPR operating limit must be modified as required by c
Specification 3.5.K.
L D.
At all reactor operating pressures, a Once a shift, check the status of the rud accumulator may be inoperable pressure and level alarms for each provided that no other control rod in accumulator.
the nine-rod square array around that rod has:
L 1.
(-
3.3/4.3-7 Amendment No. 114
a.1
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-QUAD-CITIES
.DPR-29 6.
The operability of the Scram Discharge Volume vent and drain valves assures lthe proper venting and draining of. the Volume, so that water accumulation in the Volume'does not occur.
These specifications provide for the periodic verification that the valves. are open, and for.the testing of these valves under reactor scram conditions:
during each Refueling Outage.
C.
Scram Insertion Times 1
The control rod system is analyzed to bring the reactor subcritical at a rate-fast enough to prevent fuel damage,. i.e., to prevent the MCPR from becoming less than the fuel cladding integrity safety. limit.
Ana' lysis of the. limiting power transient shows 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 required protection, and MCPR remains 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 reache's se h4 spe6N f.70;;;er.deonacyclecumulativebasis(overal1~ average'of w h coat surveillance data-te date) in order to comply with assumptions in the implementation procedure for the 00YN transient analysis computer code.
onaAtwo umTs The basis for. choosing 0.70 ;e;;r.demis discussed further in the bases
.Rarort-for. Specification'3.5.K.
In the analytical treatment of the transients, P 290 milliseconds are allowed between a neutron sensor: reaching the. scram Es sof, point and the start of motion of the control rods.
This is adequate and 7,1 conservative when compared to the typically observed time delay of about %w 210 milliseconds.
Approximately 90 milliseconds after neutron flux reaches the trip point, the' pilot scram valve solanoid 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 allowable scram insertion times specified in Specification 3.3.C.
The scram: times for all control rods will be determined at the time of each refueling ~ outage.
A representative sample of control' rods will be scram tested during the interval of greater than 16 weeks but not more than 32 weeks.
I 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 of drive performance accumulated to date indicates that the 90% insertion times of new and overhauled drives approximate a normal distribution about the mean which tends to become skewed tev. rd longer 3.3/4.3-14 Amendment No. 114
_-_I__--------_----_------
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QUAD-CITIES i
DPR-29 l
I.
Average Planar LHGR I.
Average Planar LHGR During steady-state power operation, Daily during steady-state operation i
the average linear heat generation above 25% rated thermal power, the rate (APLHGR) of all the rods in any average planar LHGR shall be deter-fuel assembly, as a function of aver-mined.
age planar exposure, at any axial lo-cation, shall not exceed the maximum average planar LHGR ;h;w. i, I';;r; specified in 6e CORE
- 0. '., L If at any time during opera-orERATigs Lin Ts REPonT.
tion it is determined by normal sur-veillance that the limiting value for APLHGR is being exceeded, action shall be initiated within 15 minutes i
to restore operation to within the I
prescribed limits.
If the APLHGR is not returned to within the prescribed limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, the reactor shall be brought to 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 />.
Surveil-lar,ce and corresponding action shall continue until reactor operation is within the prescribed limits.
J.
Local LHGR J.
Local LHGR During steady-state power operation, Daily'during steady-state power the linear beat generation rate operation above 25% of rated thermal (LHGR) of any rod in any fuel assem-power, the local LHGR shall be bly at any axial location shall not determined.
exceed the maximum allowable LHGR :
f If at any time during operation it is aetermined by normal surveillance that the limiting value for LHGR is speciCe4 la %e CcRE being exceeded, action shall be oPERATsWG LIMITS REPORT.
initiated within 15 minutes to restore operation to within the prescribed limits.
If the LHGR is not returned to within the prescribed limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, the reactor shall be brought to 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.
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.,.,m 3.5/4.5-13 Amendment No. 114
QUAD-CITIES DPa-29 K.
Minimum Critical Power Ratio (MCPR)
K.
Minimum Critical Power Ratio (MCPR)
During steady-state operation at i
rated core flow, MCPR shall be The MCPR shall be determined daily during -!
steady-state power operation above 25% of i
gr::t r th:r. Or : ;u:1 t::
rated thermal power.
1.33 for t 1 0.71 sec AVE f
eg 4.., 3,.a., - m een j
limit 5ftd{(ed in +be CORE cPERATt40
(
1.37 rt 1 0.86 sec AVE uruTs REfoRT.
0.278 tAVE +.131-for 0.71 see 1 AVE < 0.86 see where t mean 0% scram AV insert n time for all surv 11ance 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 Wa, ir, fig.;r: 3. 5.g If any time during operation it is determined by normal surveillance that the limiting value for MCPR is specni=4 la *e EoRE being exceeded, action shall be openATids units REeoRT.
initiated within 15 minutes to restore operation to within the prescribed limits.
If the steady-state MCPR is not returned to within the prescribed limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, the reactor shall be brought to 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.
3.5/4.5-14 Amendment No. 114
. [Q v
iQ QUAD-C! TIES
.a
-DPR-29 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 affact the calculated peak cladding. temperature by less.
than t20*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.t.; r. ir, ri pre 0.",-1 are baned on calculations emolovina
~
the models described in Reference 2.\\
eciOed it we cesE OPERATwo LIMITS REPost The Average Planar Linear Heat Generation Rate (APLHGR) also serves-a.
secbndary function which is to assure fuel rod mechanical; integrity.
J.
Local LHGRz Sf*ci6ed la he coat optaATmr. unirs atroar This specification assures.that the maximum linear heat generation rate in any rod is less than the design linear heat generation rate dven if fuel pellet densification is postulated.
The power spike penalty is discussed in Reference 2 and asu mts a linearly increasing variation in axial gapa between core bottom all top.and assures.with 95% confidence that no more than'one fuel rod exceeds the. design.LHGR due to power spiking.
No penalty is r? quired in Specification 3.5.L because it has been accounted for in the reload transient analyses by. increasing the calculated peak LHGR by.2.2%.
A. coat oesgAmo tinirs L
Minimum Critical Power Ratio (MCPR)
REPoaY Thesteadystate~valuesfor.MCPRspecifiedintI!;;;;;if'i;; tier,were selected to provide margin to accommodate 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 initial condition assumed for the LOCA analysis plus two percent for uncertainty 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 operatir.g 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 tr.;
- .;ificetterJor the limiting condition of operation bounds the initial value of MCPR assumed to exist prior to the initiation of the transients.
This initial conditior, which is used in the transient analyses, will preclude violation or t N fuel cladding integrity safety limit.
Assumptions and methods used in calculating the required steady state 4
MCPR limit for each reload cycle are documented in References 2 and 4.
The results apply with increased conservatism while operating with MCPRs greater than specified.
l 4.
coat oPtaATw o tinirs garoar 3.5/4.5-20 Amendment No. 114
-)
1
E-o OVAD-CIT!ES DPR-29 Operation of the fans and coolers is required during shutdown and thus additional surveillance is not required.
Verification that access doors to each vault are closed following entrance by personnel is covered by station operating procedures.
The LHGR shall be checked daily to determine if fuel burnup'or control rod i
movement has caused changes in power distribution.
Since changes due to burnup are slow and only a few control rods are moved daily, a daily check of power distribution is adequate.
Average Planar LHGR At core thermhl power levels less than or equal to 25%, operating plant experience and thermal hydraulic analyses indicate that the resulting average planar LHGR is below the maximum average planar LHGR by a considerable margin; therefore, evaluation of the average planar LHGR below this power level is not necessary.
The daily requirement for calculating average planar LHGR above 25%
rated thermal power is sufficient, since power distribution shifts are slow when there have not been significant power or control rod changes.
Local LHGR The LHGR as a function of core height shall be checked daily during reactor operation at greater than or equal to 25% power to determine if fuel burnup or control rod movement has caused changes in power distribution.
A limiting LHGR value is precluded by considerable margin when employing any permissible control rod pattern below 25% rated thermal power.
Minimum C;iti::a1 Power Ratio (MCPR)
At core thermal power levels less than or equal to 25%, the reactor will be operating at minimum recirculation pump speed and the moderator void content will be very small.
For all designated control rod patterns which may be employed at i
this point, operating plant experience and therral hydraulic analysis indicate that the resulting MCPR value is in excess of requirements by a considerable l
margin.
With this low void content any inadvertent core flow increase would l
only place operation in a more conse,rvative mode relative to MCPR.
The daily requirement for calculating MPR above 25% rated therwal power is sufficient, since-power distribution 'sif fts are very slow when there have not been significant power or control rod changes.
In addition, the K correction :
7 applied to the LCO provides margin for flow increases from low flows.
, as specJicJ in h CORE O PERATNG UMITS REPoRTy t
3.5/4.5-26 Amendment No.124
-QUAO'CsTIES OPR-29
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Amendment No. 114 l
QUA0 CITIES CPR-29
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QUA0 CITIES
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QUAD CITIES
'J OPR-29 4.
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DELETE THIS PAGE Amendment No.114
QUAD CITIES.
OPR-29 4
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Figure 3.5-1 (Sheet 5 of 5)
- " "d" "I " "
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QUAD-CITIES 11 DPR-29 y
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l DELETE THis PAGE FIGURE 3.5-2 Amendment No. 114 K FACTOR g
j
4
.(
i QUAD-CITfES DPR-29 i
2.
If Specification 3.6.H.1 cannot l
be met, one recirculation pump j
shall be tripped.
t 3.
During Single Loop Operation for more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, the following restrictions are required:
a.
The MCPR Safety Limit shall be increased by 0.01 (T.S.
1.1A);
b.
The MCPR Operating Limit :
,es spe6fieJ in we cent shall be increased by 0.01 eecanado untTs REroRT (T.S. 3.5.K);
3 c.
The f 7ow biased APRM Scram and Rad Block Setpoints shall be reduced by 3.5% to read as follows:
T.S. 2.1.A.1; 5 1 3WD + 58.5 T.S. 2.1.A.1;*
S 1 (.58WD + 58.5) FRP/MFLPD T.S. 2.1.B; 5 1 58WD + 46.5 T.S. 2.1.B;*
S 1 (.58WD + 46.5) FRP/MFLPD T.S. 3.2.C (Table 3.2-3);*
APRM Upscale < (.58WD + 46.5)
FRP/MFLPD o
d.
The flow biased RBM Rod Block setpointsahall be
- as specified in ne core reduced by 4.0%.:: r::: ;;-
ortgATigo units REPoC; T.O. 0.2.0 'TeLie 3.2 0),
G 4ecei i.0'7#-
00 e.
The suction valve in the idle loop shall be closed and electrically isol id except when the idle loop is being prepared for return to service.
3.6/4.6-11 Amendment No. 114
c QUAD-CITIES OPR-29 H.-
Recirculation Pump Flow Limitations The LPCI loop selection logic is described in the SAR, Section 6.2.4.2.5.
For some limited low probability accidents with the recirculation loop operating with large speed differences, it is possible for the. logic to select the wrong loop for injection.
For these limited conditions, the core spray itself is adequate to prevent fuel temperatures from exceeding allowable limits.
However, to limit the probability even further, a procedural limitation has been placed on the allowable variation in speeit between the recirculation pumps.
The licensee's analyses indicate that above 80% power the loop select logic could not be expected to function at a speed differential of 15%.
Below 80%
power, the loop select logic would not be expected to function at a speed differential of 20%.
This specification provides a margin of 5% in pump speed differential before a problem could arise.
If the reactor is operating on one pump, the loop select logic trips that pump before making the loop selection.
Analyses have been performed which support indefinite single loop operation provided the appropriate restrictions are implemented within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
The MCPR Safety Limit has been increased by 0.01 to account for core flow and TIP reading uncertainties which are used in the statistical analysis of the safety limit.
The MCPR Operatina limit,has also been increased by 0.01 to maintain the same margin to the safety limit as during Dual Loop operation.
The flow biased scram and rod block setpoints are reduced to account for uncertainties associated with backflow through the idle jet pumps when the operating recirculation pump is above 20-40% of rated speed.
This assures that the flow biased trips and blocks occur at conservative neutron flux levels for a given core flow.
The closure of the suction valve in the idle loop prevents the loss of LPCI flow through the idle recirculation pump into the downcomer.
I.
Snubbers All snubbers are required OPERABLE to ensure that the structural integrity of the reactor coolant system and all other safety-related systems is maintained I
during and following a seismic or other event initiating dynamic loads.
Snubbers excluded from this inspection program are those installed on non-safety-related systems and then only if their failure or failure of the system on which they are installed, would have no adverse effect on any safety-related system.
The visual inspection frequency is based upon maintaining a constant level of snubber protection to systems.
Therefore, the required inspection interval I
varies inversely with the observed snubber failures and is determined by the number of inoperable snubbers found during an inspection.
Inspections performed before that interval has elapsed may be used as a new reference point to determine the next inspection.
However, the results of such early inspections performed before the original required time interval has elapsed (nominal time less 25%) may not be used to lengthen the required inspection interval.
Any inspection whose results require a shorter inspection interval will override the previous schedule.
, e creemed ia we. cott i
CPERATolG LtHtTs EEPoRT, 3.6/4.6-24 Amendment No. 114
.c*..
~
QUAD-CITfES DPR-29 whole body dose received from external sources shall be ass'igned to specific major work functions.
3.
Monthly Operating Report Routine reports of operating statistics and shutdown experience.
shall. be submitted on a monthly basis to the Director, Office of..
Management Information and. Program Control, U.S. Nuclear Regulatory Commission, Washington, DC 20555, with a copy to the-i appropriate Regional Office, to arrive no later than the 15th.of each month following the calendar-' month covered by the report.
In addition, any changes to the'0DCM shall be submitted with the Monthly Operating Report within 90 days of the effective date of the change.
A report:of major change to the radioactive. waste treatment systems shall be submitted with the Monthly Operating Report for the period in which the evaluation was reviewed and accepted by the onsite review function.
If such change is re-evaluated and-INSERT not. installed, notification of cancellation of the change should g
be provided to the NRC.
B.
Unique Reporting Requirements 1.
Radioactive Effluent Release Report (Semi-Annual)
A semi-annual report shall be submitted to the Commission within 60 days after January 1 and July 1 of each year specifying the quantity of each of the radionuclides released to unrestricted
^
areas in liquid and gaseous effluents during the previous 6 months.
The format and content of the report shall be in accordance with Regulatory Guide 1.21 (Revision 1) dated June, 1974.
Any changes to the PCP shall be included in this report.
2.
Environmental' Program Data (Annual Report)
An annual report containing the data taken in the standard radiological monitoring program (Table 4.8-4) shall be submitted prior to May 1 of each year.
The content of the report shall include:
a.
Results of all environmental measurements summarized in the j
format of the Regulatory Guide 4.8 Table 1 (December 1975).
)
(Individual sample results will be retained at the Station).
j In the event that some results are not available for inclusion I
with the report, the report shall ne submitted noting and l
explaining the reasons for the missing results.
Summaries,
]
l interpretations, and analysis of trends of the results are to q
be provided.
I 6.6-2 Amendment No. 114 a
INSERT (for DPR-29) 4.
Core Operating Limits Report Core operating limits shall be established and documented in the a.
CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle.
b.
The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by NRC in NEDE-24011-P-A, General Electric Standard Application for Reactor Fuel (latest approved revision).
c.
The core operating limits shall be determined so that all applicable limits (e.g.,
fuel thermal-mechanical limits, core 1
thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.
d.
The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the US NRC Document Control
?esk with copies to the Regional Administrator and Resident Inbpector.
l l
0188T24
QUAD CITIES-OPR-30 TECHNICAL SPECIFICATIONS APPENDIX A LIST OF FIGURES Number Title 2.1-1 APRM Flow Reference Scram and APRM Rod Block Settings 2.1 Deleted 2.1-3 APRM Flow Bias Scram Relationship to Normal _ Operating Conditions 4.1-1 Graphical Aid in the Selection of and Adequate Interval Between Tests 4.2-1 Test Interval vs. System Unavailability 3.4-1 Deleted 3.4-2 Sodium Pentaborate Solution Temperature Requirements 3.5-1 Deleted 3.5-2 Deleted 3.6-1 Minimum Temperature Requirements per Appendix G of 10 CFR 50 4.6-1 Chloride Stress Corrosion Test Results at 500*F 4.8-1 Locations of Fixed Environmental Radiological Monitoring Stress 6.1-1 Deleted 6.1-2 Deleted 6.1-3 Minimum Shift Manning Chart I
l 1976H vi Amendment No.
d"
.?
QUADLCITIES
?
.DPR.
1
~ DEFINITIONS s
The succeeding frequently used terms are explicitly-defined.so that a uniform.
.in_terpretation of the_ specifications may be achieved.
A.
Alteration ~of the Reactor Core - The act of moving any. component.in the region above the core support-plate, below the upper-grid, and within the
- shroud.
Normal control rod movement with the control rod drive hydraulic' system is not defined as-a core: alteration.
Normal movement of'incore:
. Instrumentation or movement of the TIP system is not defined as a core alteration.
.B.
CORE OPERATING LIMITS REPORT - The unit specific document that provides core operating limits for the current operating reload cycle.
These cycle specific core operating limits shall be determined for each reload
. cycle in accordance with Specification 6.6.
Plant operations within these operating limits'is addressed in individual specifications.
C.
Hot Standby'- Hot standby means operation with the reactor-critical, systein pressure less than 1060 psig, the main steam isolation valves closed, and thermal power not exceeding 15%.
D.'
Immediate --Immediate means that the required action will be initiated as soon as practicable considering the safe operation of the unit and the importance of the required action.
E.
Instrument Calibration - An instrument calibration means the adjustment of an instrument signal output so that it corresponds, within acceptable range and accuracy, to a known value (values) of the parameter wh.ich the instrument, monitors. Calibration shall encompass the entire instrument, including actuation, alarm, or trip.
Response time is not part.of the routine. Instrument calibration but will be checked once per operating cycle.
F.
Instrument Functional Test - An instrument functional test means the injection of:a simulated signal into the instrument primary sensor to verify the proper instrument response alarm and/or initiating action.
'G.
. Instrument Check - An instrument check is qualitative determination of acceptable operability by observation of instrument behavior during operation.
This determination shall include, where possible, comparison of the instrument with other independent instruments measuring the same variable. -
L l
i 1976H I.0-1 Amendment No.
L_-
~
QUAD C8T2ES DPR-30'
.l
, 1.1 SAFE 7Y LfMIT BASIS The fuel cladding integrity limit is set such that no calculated fuel damage would occur as a result of an abnornal operational transient. Because fuel danage is not directly observable, a step-back approach is used to establish a safety limit such that the minimum critical power ratio (MCPR) is no less than the fuel cladding integrity safety limit MCPR
> the fuel cladding integrtty safety limit represents a conservative margin relative to the conditions required to maintain fuel cladding integrity.
The fuel cladding is one of the physical boundaries which separate radioactive materials from the environs. The integrity of the fuel cladding is related tw its relative freedom from perforations or cracking.
3 Although some corrosion or use-related cracking may occur during the life of the cladding.
fission product migration from this source is incrementally cumulative and continuously measurable. - Fuel cladding perf orations, however, can result f rom thernal stresses which occur f rom reactor operation significantly above design conditions and the protection system safety settings. While fission product migration from cladding perforation is just as measurable as that from use-related cracking, the thernally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration. Therefore, the fuel cladding safety limit is defined with margin to the conditions which would produce onset of transition boiling (MCPR of 1.0).
These conditions represent a significant departure fran the condition intended by design for planned operation. Therefore, the fuel cladding integrity safety limit is established such that no calculated fuel damage shall result from an abnormal operational transient. Basis of the values derived for this safety limit for each fuel type is documented in References 1 and 2.
A.
Reactor Pressure > 800 psig and Core Flow > 10% of Rated Onset of transition boiling results in a decrease in heat transfer from the cladding and therefore elevated cladding temperature and the possibility of cladding failure. However, the existence of critical power. or boiling transition is not a directly observable parameter in an operating reactor.
Therefore, the margin to boiling transition is calculated from plant operating parameters such as core power. core flow, feedwater temperature, and core power distribution. The margin for each fuel assembly is characterized by the critical power ratio (CPR). which is the ratto for the bundle power which would produce onset of transition boiling divided by the actual bundle power. The minimum value of this ratio for any bundle in the core is the minimum critical power ratio (MCPR). It is assumed that the plant operation is controlled to the nominal protective setpoints via the instrumented variables (Figure 2.1-3).
The MCPR fuel cladding integrity safety limit has sufficient conservatism to assure that in the event of an abnornal operational transient initiated from the normal operation condition, more than 99.9% of the fuel rods in the core are expected t3 avoid boiling transition. The margin between MCPR of 1.0 (onset of transition boiling) and the safety limit. is derived from a detailed statistical analysis considering all of the uncertainties in monitoring the core operating state, including uncertainty in the boiling trans1 tion correlation (see e.g..
Reference 1).
Because the boiling transition correlation is based on a large quantity of full-scale data, there is a very high confidence that operation of a fuel assenbly at the condition of MCPR - the fuel cladding integrity safety limit would not produce boiling transition.
However, if boiling transition were to occur cladding perforation would not. be expected. Cladding temperature would increase to approximately 1100*F, which is below the perforation temperature of the cladding material. This had been verified by tests in the General Electric Test Reactor (GETR), where similar l
fuel operated above the critical heat flux for a significant period of time (30 minutes) without cladding perforation.
If reactor pressure should ever exceed 1400 psia during normal power operation l
(the limit of applicability of the boiling transition correlation), it would bc l
assumed that the fuel cladding integrity safety limit has been violated.
In addition to the boiltng transition limit (MCPR) operation 1s constrained to a maximum LHCR specified in the CORE OPERATING LIMITS REPORT for various fuel types. This constraint is established by Specification 3.5.J. to provide adequate safety margin to 1% plastic strain for abnormal operating transients initiated f rom high power conditions. Specification 2.1.A.1 provides for equivalent safety margin for transients initiated f rom lower power conditions by adjusting the APRM flow-biased scram setting by the ratto of FRP/MFLPD.
I 1977H 1.1/2.1-4 Amendnent No.
E l'.
f QUAD CITIES.
~
~
j 2.1 LIMITING SAFETY SYSTEM SETTING BASES l
The abnormal operational. transients applicable to. operation of the units have been analyzed throughout the spectrum of planned 6perating conditions in accordance with Regulatory Guide 1.49.
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.
Conservatism incorporated into the transient analysis is documented in References 1 and 2.
Transient analyses'are initiated at the conditions given in these References.
The scram delay time and rate of rod insertion allowed by-the analyses are L
conservatively set equal to the longest delay and slowest insertion rate I
acceptable by technical specifications.
The effects of scram worth, scram delay l
time, and rod insertion rate, all conservatively applied, are of greatest j
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.
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 the CORE OPERATING LIMITS REPORT as the limiting condition of operation bound l
those which are conservatively assumed to exist prior to initiation of the transients.
A.
Neutron Flux Trip Settings 1.
APRM Flux Scram trip Set.ing (Run Mode)
The average power range monitoring (APRM) Gystem, 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.
l 0926B/03982 1.1/2.1-7 Ame.dment No.
E-q q
(J c,
QUAD-CITIES y,
DPR-30 i
{
'.. j TABLE 3.2-3 INSTRUMENTATION THAT INITIATES ROD BLOCK
~ Minimum Number.
of Operable or Tripped Instrument.
Channels per Trio System U2.
Instrument' Trio Level Setting 2-
- APRM upscale (flow blas)I73 1[0.58HD +l50]
FRP [2]'
MFLPD 2
APRM upscale (Refuel and 112/125 full. scale
- Startup/ Hot Standby mode) 2 APRM downscaleI73 23/125 full scale 1
Rod block monitor upscale (flow
[10]
.blas)[73 1
Rod block monitor downscaleI73 23/125 full scale 3-IRM downscale[3] [8]
23/125 full scale 3
IRM up' scale [8]
1 08/125 full scale 1
2[5]
.SRM detector n t in Startup 22 feet below core centerline position I4 3
IRH detector,nQt in Startup 22 feet below core centerline position 68J 2[5] [6]
SRM upscale 1105 counts /sec 2[5]
SRM downscale l93 1102 counts /sec-1 (per bank) High water level in scram 1 25 gallons (per bank)
-discharge volume (SDV) 1-SDV high water level scr.am NA trip bypassed Notes
.l.
For the Startup/ Hot Standby and Run positions of the reactor mode selector switch, there shall be two operable or tripped trip systems for each function except the SRM rod blocks.
IRM upscale and IRM downscale need not be operable in the'Run position. APRM downscale, APRM upscale (flow biased), and RBM downscale need not be operable in the Startup/ Hot Standby mode.
The RBM upscale need not be operable at less than 30% rated thermal power. One channel may be bypassed above 30% rated thermal power provided that a limiting control rod pattern does not exist.
For systems with more than one channel per trip system, if the first column cannot be met for one of the two trip systems, this condition may exist for up to 7 days provided that during that time the operable system is functionally tested immediately and daily thereafter; if this condition lasts longer than 7 days the system shall be tripped.
If the first column cannot be met for both trip systems, the systems shall be tripped.
1976H/0615Z 3.2/4.2-14 Amendment No.
9
._m._
QUAD-CITIES DPR-30 2.
Hp is the percent of drive flow required to produce a rated core flow of 98 million Ib/hr. Trip level setting is in percent of rated power (2511 MHt).
3.
IRM downscale may be bypassed when it is on its lowest range.
4.
This function is bypassed when the count rate is > 100 cps.
5.
One of the four SRM inputs may be bypassed.
6.
This SRM function may be bypassed in the high IRM ranges (ranges 8, 9, and 10) when the IRM upscale rod block is operable.
7.
Not required to be 'perable when performing low power physics tests at atmospheric pressure during or after refueling at power levels not to exceed 5 MWt.
8.
This IRM function occurs when the reactor mode switch is in the Refuel or Startup/ Hot Standby position.
9.
This trip is bypassed when the SRM is fully inserted.
- 10. The Rod Block Monitor upscale setpcint shall be established as specified in the CORE OPERATING LIMITS REPORT.
1976H 3.2/4.2-14a Amendment No.
1
-.?
QUAD-CITIES i
OPR-30 l
sidered inoperable, fully proviGe reasonable assurance
)
inserted into the core, and that proper control rod drive l
electrically disarmed.
performance is being j
maintained. The results of j
5.
If the overall average of the measurements performed on the 20% insertion scram time data control rod drives shall be generated to date in the current submitted in the annual cycle exceeds the limit operating report to the NRC.
specified in the CORE OPERATING i
LIMITS REPORT, the MCPR 5.
The cycle cumulative mean scram 1
operating limit must be modified tine for 20% insertion will be as required by Specification determiced innediately following 3.5.K.
the testing required in Specifications 4.3.C.1 and 4.3.C.2 and the MCPR operating limit adjusted, if necessary, as required by Specification 3.5.K.
D.
Control Rod Accumulators At all reactor operating pressures, a Once a shift. check the status of the rod accumulator may be inoperable pressure and level alarms for each provided that no other r.ontrol rod in accumulator.
j the nine-rod square array around that i
rod has:
1.
An inoperable accumulator, 2.
A directional control valve electrically disarmed while in a nonfully inserted position. or 3.
A scram insertion greater than maximum permissible insertion time.
If a control rod with an inoperable accumulator is inserted full-in and its directional control valves are electrically disarred it shall not be considered to have an inoperable accumulates, and the rod block asso-ciated with that inoperable accumu-lator may te bypassed.
E.
Reactivity Anomalies E.
Reactivity Anomalies The reactivity equivalent of the dif-During the startup test program and ference between the actual crit 1 cal startups following refueling outages.
rod configuration and the expected the critical rod configurations will configuration during power operation be compared to the expected configur-shall not exceed 1% 4 k.
If this ations at selected operating condi-limit is exceeded, the reactor shall tions. These comparisons will be be shutdown until the cause has been used as base data for reactivity determined and corrective actions monitoring during subsequent power have been taken. In accordance with operation throughout the fuel cycle.
Specif'. cation 6.6. the NRC shall be At specific power operating condi-notified of this reportable occur-tions, the critical rod configuration rence within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, will be compared to the configurat n expected based upon appropriately corrected past data. This comparison will be made at least every equiva-lent full power month, f.
Economic Generation Control System F.
Economic Generation Control System Operation of the unit with the eco-Prior to entering EGC and once per nomic generation control system with shift while operating in EGC. the EGC automatic flow control shall be per-operating parameters will be reviewed missible only in the range of 65% to for acceptability.
100% of rated core flow. with reactor power above 20%.
1977H 3.3/4.3-5 Amendnent No.
' QUAD CITIES m
CPP.-30
+,'
C..
Scram Insertion Times The control rod system is analyzed to bring the reactor subtritical at a rate fast enough to prevent fuel damage, i.e., to prevent the MCPR from becoming less.tban the fuel cladding integrity safety limit.
Analysis of.the limiting power transient shows.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 required protection, and MCPR remains' 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 the limit specified in the CORE OPERATING LIMITS REPORT 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 this 20% scram insertion' time. limit is discussed further in l the bases for Specification 3.5.K.
In the analytical treatment of the tiansients, 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 contrc' 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 allowable scram insertion times specified in Specification 3.3.C.
The scram' times for all-control rods will be determined at the time of each refueling outage. A representative sample of control rods will be scram tested during the interval of greater than 16 weeks but not more than 32 weeks.
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 of drive performance accumulated to date indicates that the 90%
insertion times of new and overhauled drives approximate a normal distribution about the mean which tends to become skewed toward longer scram times 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 performance will detect local variations and also provide assurance that local scram time limits are not exceeded. Continaed monitoring of other drives exceeding the expected range cf scram times provides surveillance of possible anomalous performance.
The numerical values assigned to the predicted scram performance are based on the analysis of the Dresden 2 startup data and of data from other BWR's such as Nine Mile Point and Oyster Creek The occurrence of scram times within the limits, but significantly longer than average, should be viewed as an indication of a systematic problem with control rod drives, especially if the number of drives exhibiting such scram times exceeds eight, the allowable number of inoperable rods.
1976H 3.3/4.3-10 Amendment No.
L
3 QUAD-CITIES.
DPR-30 j
j 3.
If Sa cification 3.5.H.1 and 2 b.
During each operating cycle, j
cannot be met, reactor startup the following flood protection 1
shall not commence or if level switches shall be i
operating an orderly shutdown functionally tested to give the shall be initiated and-the following control room alarms:
reactor shall be in a cold shutdown condition within 24 1)
. turbine building equipment hours.
Pain sump high level 2) vault high level c.
The RHR service water vault sump pump discharge check valves outside the vault shall be tested for integrity, using clean demineralized water, at least once per operating cycle.
d.
The condenser pit 5-foot crip circuits for each channel shall be checked once a month. A logic system functional test shall be performed during each refueling outage.
I.
Average Planar LHGR I.
Average Planar LHGR During steady-state power operation, Daily during steady-state operation the average linear heat generation above 257. rated thermal power, the rate (APLHGR) of all the rods in any average planar LHGR shall be deter-fuel assembly, as a function of mined.
average planar exposure, at any axial location, shall not exceed the maximum average planar LHGR specified in the CORE OPERATING LIMITS REPORT.
If at any time during operation it is determined by normal surveillance that the limiting value for APLHGR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits.
If the APLHGR is not returned in 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 re ctor operation is within the prescribed limits.
1976H 3.5/4.5-9 Amendment No.
1
[
e.
. QUAD-CITIES DPR-30 l - c, e
N 4
.J.
Local LHGR' J.
Local LHGR Durin'g stesdy-state power operation',
Daily during steady-state power..
the linear heat. generation rate operation above'25% of rated thermal (LHGR) of any rod in any fuel power, the local LHGR.shall be assembly at any axial.' location shall determined.
not exceed the maximum allowable LHGR-specified.in the CORE OPERATING LIMITS REPORT.
If at any time during operation it is determined by normal surveillance that the limiting value for LHGR is'being exceeded, action shall be. initiated within 15 minutes to restore operation to within the prescribed limits, If the LHGR is not returned.to within'the prescribed limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, the reactor
.shall be brought to 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 shallicontinue until reactor operation is within the-prescribed limits.
K.
Minimum Critical Power Ratio-(MCPR)
K.
Minimum Critical Power Ratio.(MCPR)
During. steady-state operation at
.The MCPR shall be determined.dailj.
rated core flow', MCPR '; hall be equal during steady-state power operation to or greater than.the MCPR limit above 25% of' rated. thermal power, specified in the CORE OPERATING
. LIMITS REPORT.
'For core' flows other than rated,
'these nominal values of MCPR shall be increased by a factor of kr where kf is as specified in the CORE OPERATING LIMITS REPORT.
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 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 untti reactor operation is w M in the
. prescribed limits.
.1976H 3.5/4.5-10 Amendment No.
F'
., e c 00AD-CITIES
?
- '*/
DPR-30 l.'.
i{)
' diesel generators.' All cf:these systems h'av2 bien sized to p2rfcrm their intended function considering the simultaneous operation of both units.
These technical specifications contain only'a single reference to the operability and surveillance requirements for the shared safety-related features of each plant. -The level of operability for one unit must be
' maintained independently of the status of the other. For. example a diesel (1/2 diesel) whicts is shared between Units 1 and 2 would have to be operable for continuing Unit 1 operation even if Unit 2 were in a cold shutdown condition and needed no diesel power.
Specification 3.5.F.3 'provides that should this occur, no work will be performed which could preclude adequate emergency cooling capability being available. Work is prohibited unless it is in accordance with specified procedures which limit the period that the control rod drive housing is open and assures that the worst possible loss of coolant resulting from the work will not result in uncovering the reactor core. 'Thus, this specification assures adequate core cooling. Specification 3.9 must be consulted to determine other-requirements for the diesel generator, i
G.
' Maintenance of Filled Discharge Pipe -
1
'If the discharge piping of the core spray, LPCI mode of the RHR. HPCI.
and RCIC are not filled, a water hanener can develop in this piping, threatening system damage and thus the availability of emergency cooling.
systems when the pump and/or pumps are started. An analysis has been done which shows that if a water hanener were to occur at the time emergency cooling was required, the systems would still perform their design function. However to minimize damage to the discharge systems and to ensure added margin in the operation of these systems, this technical specification requires the discharge lines to be filled whenever th9 system is in an operable condition.
Specification 3.3.F.4 provides assurance that an adequato supply of coolant water is insnediately available to the low-pressure core cooling systems and that the core will remain covered in the event of a loss-of-coolant accident while the reactor is depressurized with the head removed.
H.-
Condensate Pump Room Flood Protection See Specification 3.5.H 1.
Average Planar LHGR This specification assures that the peak cladding temperature following a postulated design-basis loss-of-coolant accident will not exceed the 2200'F limit' spec 1t ied in 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 LHGR of all the. rods in a fuel assembly at any axial location and is only secondarily dependent on the rod-to-rod power distribution within a fuel assembly. Since expected local varlations in power distribution within a fuel assembly affect the calculated peak cladding temperature by less than 120*F relative to the paak temperature for a typical fuel design, the limit on the average planar LHGR is suf ficient to assure that calculated temperatures are below the 10 CFR 50 Appendix K limit.
The maximum average planar LHGR's specified in f.he CORE OPERATING LIMITS REPORT are based on calculations employing the models described in Reference 2.
Power operation with LHGR's at or below those specified in the LORE OPERATING LIMITS REPORT assures that the peak cladding l
temperature following a postulated loss-of-coolant accident will not exceed the 2200*F limit. These values represent limits for operaticn to ensure conformance with 10 CFR 50 and Appendix K only if they are more limiting than other design parameters.
The maximum average planar LHGR's specified in the CORE OPERATING LIMITS l
REPORT at higher exposures result in a peak cladding temperature of less I
than 2200*F. However, the maximum average planar LHGR's are it 77H 3.5/4.5-13 Amendment No.
l
~
QUAD-CITIES DfR-30 specified in the CORE OPERATI%G LIMITS REPORT as limits becausa confermanca l
.g#
calculations.have n t b un p2rformed to justify op:ratign at LHGR's in exC2ss of those shown.
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 specified in the CORE OPERATING LIMITS REPORT 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 assures with 95% confidence that no more than one fuel rod exceeds the design LNGR due to power spiking. No penalty is tequired in Specification 3.5.L because it has been accounted for in the reload l transient analyses by increasing the calculated peak LHGR by 2.2%.
K.
Minimum CritiO I Power Ratio (MCPR)
'The steady state values for MCPR specified in the CORE OPERATING LIMITS REPORT were l
selected to provide margin to acconsnodate 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 initial condition assumed for the LM A analysis plus two percent for uncertainty is satisfied. For any of the special set of transients or disturbances caused by single operator error or singlo 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 transientt, the value of MCPR stated in the CORE OPERATING LIMITS REPORT for the limiting condition of operation bounds the initial value of MCPR assumed to exist prior to the initiation of the transients. This initial condition, which.15 used in the transient analyses, will preclude violation
'of the 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 and 4.
The 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
~
c)
Loss of feedwater heater Several f actors influence which of 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 Limiting Condition of Operation are conservatively chosen to bound the most restrictive over the entire cycle for each fuel type.
The need to adjust the MCPR operating limit as a function of scraia time arises from the statistical approach used in the implementation of the ODYN computer code for analyzing rapid pressurization events. Generic statistical analyses were performed for plant groupings of similar design which considered the statistical variation in several para'neters (initial power level. CR0 scram insertion time, and model uncertainty). These analyses (which are described further in Reference 4) produced generic Statistical Adjustment Factors which have been applied to plSnt and cycle specific ODYN results to yield operating limits which provide a 95% cobability with 95% confidence that the limiting pressurization event will not cause MCPR to fall below the fuel cladding integrity safety limit.
l 1977H/0615Z 3.5/4.5-14 Amendment No.
i
___-____-_____O
i QUAD-CITIES-a 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 occurring 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.C.I.
In practice, however, the requirements of 3.3.C.1 would most likely be reached (i.e. Individual data set average >.90 secs) and the required act %ns taken (3.3.C.2) well before the running average exceeds 0.90 secs.
The 5% significance level is defined in Reference 4 as:
n TB - p + 1.65 (N / I Nj)1/2 a 1
1-1 where:
Mean value for statistical scram time distribution.to 20%
p inserted standard deviation of above distribution a
=
number of rods tested at BOC (all operable rods)
N1
=
n total number of operable rods tested in the current cycle I Nj 1-1 The value.for TB used in Specification 3.5 k is specified in the CORE OPERATING LIMITS REPORT and is conservative for the following reasons:
a)
For simplicity in formulating and implementing the LCO, a n
conservative value for I Nj of 708 (i.e. 4x177) was used.
1-1 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 p and a 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 linea,- MCPR penalty at a slightly lower value Tave.
This also prysuces the full 4.4%
penalty at 0.86 secs which would occur sooner than the required value of 0.90 secs.
1976H 3.5/4.5-14a Amendment No.
y
..,~ ~
QUAD-CZTIES DPR-30 1., -
e* g Local LHGR The LHGR as a function of core-height shall be checked daily during reactor-
. operation at greater than or equal to 25% power to determine if fuel burnup or control rod movement has caused c'anges in power distribution. A limiting LHGR value is precluded by considerable margin when employing any permissible _ control rod pattern below 25% rated. thermal power.
Minimum Critical Power Ratio (MCPR)
At core thermal power levels less than or equal to 25%, the reactor will De operating at niinimum recirculation pump speed and the moderator void content will-be very small.
For all designated control rod patterns which may be employed at this point, operating plant experience and thermal hydraulic' analysis indicate that the resulting MCPR.value is in excess of requirements by a considerable margin.
With this low void content, any inadvertent core flow increase would only place _ operation in a more conservative mode relative to MCPR.
The daily required for calculating MCPR above 25% rated thermal power is sufficient, since power distribution shifts are very slow when there have not been significant power or control rod changes.
In addition, the.Kr correction, as specified in the CORE OPERATING LIMITS REPORT, applied to the LCO provides margin for flow increases from low flows.
1976H 3.5/4.5-18 Amendment No.
.i.-
' QUAD-CITIES DPR-30
~
~ *
- 3. ~ Prior-to Single Loop Operation for.
more than 12. hours, the'following' restrictions are required:
The MCPR Safety Limit-shall.be a.-
'ncreased_by 0.01.
(T.S. l.1A);
.b..
The MCPR Operating Limit as specified in the' CORE OPERATING LIMITS REPORT,_shall be increased by 0.01 iT.S. 3.5.K);
c.
The flow biased APRM Scram'and
' Rod Block Setpoints'shall be-reduced by 3.57. to read,as.-
follows:
T.S.12.1.A.1;-
S 1 58HD + 58.5 T.S.'2.1.A.1;*
S 1 (.58HD + 58.5) FRP/MFLPD.
T.S. 2.1.B; S 1 58HD + 46.5 "T.S. 2.1.B;*
S 1-(.58HD + 46.5) FRP/MFLPD
-T.S.'3.2.C (Table'2.1-3);*
-APRM upscale 1 (.58HD + 46.5)
FRP/MFLPD In the event that MFLPD exceeds FRP.
d.
The flow 'blased RBM Rod Block setpoints, as specified in the CORE OPERATING LIMITS REPORT, shall-be reduced by 4.01..
e.
The suction valve in the idle loop shall be closed and electrically isolated except-when the idle loop is being l
prepared for return-to service.
l l
l 1976H 3.6/4.6-Sa Amendment No.
p j
[
j u
_n-
' QUAD-CITIES
,,'. j The licensee's analyses indicate that above 80% power.the loop select logic could not be expected to function at a speed differential _of 15L _ Below 80%
power, the loop select logic would not be expected to function at'a speed differential of 20%..This specification provides a margin of 5%-ir. pump speed-
. differential before a problem could arise.
If the reactor is operating on one pump, the loop select logic trips'that pump,before making the loop selection.
Analyses have been performed which support indefinite single loop operation provided the appropriate restrictions are implemented within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
The MCPR Safety Limit has been increased by.0.01 to account for core flow and TIP reading uncertainties which are used-in the statistical analysis of the safety limit.
The MCPR Operating Limit, as specified in the CORE OPERATING LIMITS REPORT, has l
also been increased by 0.01.to maintain the same margin.to.the safety. limit as during Dual Loop operation.
The flow biased scram and rod block setpoints are reduced to acc.ount for uncertainties associated with backflow through the idle jot puops when the operating recirculation pump is above 20 - 40% of rated speed. This assures that the flow biased trips and bic).r occur at conservative neutron flux. levels for a given core flow.
The closure of the suction valve in the idle loop prevents'the loss of LPCI flow
'through the idle recirculation pump into the downcomer.
)
l t
1976H 3.6/4.6-13a Amendment No.
i
' S.L-,
w p
su "iy QUAD-CITIES 4
. M :4,'. g (
't:
'{'
2.
A tabu'lation shall be submitted on an annual basis of the number-
~
of statio'nLutility, zand.other personnel.(including. contractors) receiving; exposures greater than 100 mrem /yr and their associated; D
. man-rem' exposure according'to work and. job: function (Notet. this tabulation supplements the requirements of Section 20.407 of 10 CFR 20),
e.g.,
reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance-(describe maintenance),
waste processing, and refueling.
The' dose assignments to various
- duty functions may be est'imates based ca pocket dosimeter, TLD, or film-badge measurements.
Small exposures totalling less than 20% of the individual total dose need not be accounted for.
In the aggregate, at least'00% of the total whole body dose received from 4
external sources shall be assigned to specific major work functions.
3.
Monthly. Operating Report Routin'e. reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the Director, Office of Management Information'and. Program Control, U.S. Nuclear' Regulatory Commission, Washington, D.C.
205'55; with a copy to the appropriate Regional.
Office,~to arrive no later than the 15th of each month following the calendar month covered by the report.
In addition, any changes to the ODCM shall. int' submitted with the Monthly Operating Report within 90 days of the effective date of the change.
.A' report-of major change to the radioactive waste treatment syst-,
shall be submitted with the Monthly Operating Report for the period-in which the: evaluation was reviewed and accepted by the onsite review function.
If such change is re-evaluated and not installed, notification of cancellation of the change should be provided to the NRC.
4.
Core Operating Limits Report Core operating limits shall be established and documented'in the a.
CORE OPERATING LIMITS REPORT before er.ch taload cycle or any remaining part of a reload cycle, b.
The analytical methods used to determine the core operating-limits shall be those previously reviewed and approved by HRC in h)
NEDE-24011-P-A, General Electric Standard Application for Reactor Fuel (latest approved revision).
c.
The core operating limits shall be determined so that all applicable limits (e.g.,
fuel thermal-mechanical limits, core thermal-hydraulle limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met, d.
The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements tttreto, shall be provided upon issuance, for each reload ~ycle, to the US NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.
6.6-2 knendment No.
0188T23
[.f ;
.6 AIIACIREKI_3 ElGlilELC6NT HAZARD EVALUATION-Commonwealth Edison proposes to amend Facility Operating License DRP-29 (Quad Cities. Unit 1) and DRP-30 (Q ed Cities Unit 2) to support the lacensing under 10CFR50.59 of Unit 1 cycle 11, Unit 2 cycle 11, and future Quad Cities operating. cycles. The purpose of.the proposed revision is to remove cycle-specific power distribution limits. Although a general
-description of the changes follows, more detailed discussion of the changes and their technical bases can be found'in Attachment 1.
' DESCRIEILCli_.QE_Atil2iDBENI_REQUESI The proposed change in Facility Operating License DRP-29 (Quad Cities Unit 1) and ORP-30 (Quad Cities Unit 2) is as follows:
A definition for-the Core Operating Limits Report is added.
The limits for Average Planar Linear Heat Generation Rate (APLHGR),
Minimum Critical Power Ratio (MCPR), Linear Heat Generation. Rate (LHGR),
and Rod Block Monitor (RBM) Upscale Setpoints are rear.oved and references to the Core Operating Limits Report are jnserted.
The explanations in the bases for the determination of the Average Planar
~
Linear Heat Generation Rate, Minimiun Critical Power Ratio, Lineae Heat Generation ~ Rate, and Rod Block Monitor Upscale Setpoints limits are revised to reference the Core Operating Limits Report. These bases sections discuss how APLHGR, MCPR, LHGR, and RBM setpoint limits values are calculated for the specific fuel types. These changes are administrative in nature and do not change the intent of the bases.
A.new administrative retforting requirement entitled, " Core Operating Limits Report", is ad6ed to the existing reporting requirements.
BASIS FOR PROPOSED NO IIDEIIICAUI_HAZAFDS CQUElDEEATION DETERMINATE.Q!i 4
Commonwealth Fe. son has evaluated the proposed Technical Specifications and determined that they do not represent a significant hazards consideration. Based on the criteria for defining a significant hazard established in 10 CFR 50.92(c):
_2 I
i 1.
The proposed mmendment does'not involve a significant increase in the probability or consequences of any accident previously evaluated. No plant protective functions are changed by this amendment. This amendment removes cycle-specific and fuel bundle type specific powec distribution limits faom the Technical Specifications and places them in a separate, controlled document entitled " Core Operating Limits Report (COLR)".
These changes are essentially administrative in nature,.as the parameters in the COLR are the same as those currently specified in the Technical Specifications. NRC approved methods will still be used to analyze reloads of NRC approved fuel types to determine the results reported in the COLR.
The surveillance requirements for these power distribution limits remain unchanged.
2.
The proposed amendment-does not create the possibility of a new or different kind of accident from any accident previously evaluated. No plant protective functions are changed by this amendment, thus the change does not create any new accident mode.
The current spectrum of reactor transients and accidents analyzed remains unchanged.
3.
The proposed amendment does not involve a significant reduction in the margin of safety. No plant functions are changed by this amendment. The change only removes cycle-specific and fuel bundle type specific power distribution limits from the Technical Specifications and incorporates these limits in the COLR.
The plant will continue to-be operated under these same power distribution limits, which will be calculated using NRC approved methods.
Based on the above discussion, Commonwealth Edison concludes that the proposed amendments do not represent a significant hazards consideration.
l 0188T:10-11 i
4 4
i f-ATIACIMIMI_4 1
EKAMPLE CORE OPERATIILG_L1111T REPORTS (COLR1 A.
QUAD CITIES STATION UNIT 1 RELOAD 9 (CYCLE 10)
B.
QUAD CITIES STATION UNIT.2 RELOAD 9 (CYCLE 10) l f
l 0188Tt12
________--______O