ML20151D081
| ML20151D081 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 07/07/1988 |
| From: | Perkins K Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20151D084 | List: |
| References | |
| NUDOCS 8807220342 | |
| Download: ML20151D081 (17) | |
Text
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p2Ho ug'o UNITED STATES g
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g NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20655
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IOWA ELECTRIC LIGHT AND POWER COMPANY CENTRAL IOWA POWER COOPERATIVE CORN BELT POWER COOPERATIVE DOCKET N0. 50-331 DUANE ARNOLD ENERGY CENTER AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 151 License No. DPR-49 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Iowa Electric Light and Power Company, et al., dated February 26, 1987, as supplemented June 1, Energy Act of 1954, as amended (the Act)quirements of the Atomic 1987, complies with the standards and re
, and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; f
l D.
The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. DPR-49 is hereby amended to read as follcws:
g72ggf$$$$p P
- (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.151, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
The license amendment is effective as of the date of issuance and shall be implemented within 30 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION e
, OA.
Kenneth E. Perkins, Director Project Directorate III-3 Division of Reactor Projects - III, IV, V and Special Projects
Attachment:
Changes to the Technical Specifications Date of Issuance: July 7, 1988
4 ATTACHMENT TO LICENSE AMENDMENT NO. 151 FACILITY OPERATING LICENSE NO. DPR-49 DOCKET NO. 50-331 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised areas are indicated by marginal lines.
Pages i
v vi vii 3.2-4 3.2-23 3.2-34 3.2-45 3.2-45a 3.4-4 3.4-5 3.4-6 3.4-8 3.4-9 y.
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,s DAEC-1
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TECHNICAL SPECIFICATIONS
-TABLE OF CONTENTS
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PAGE NO.
1.0 Definitions 1.0-1 LIMITING SAFETY SAFETY LIMITS SYSTEM SETTING
-1.1. Fuel Cladding Integrity 2.1 1.1-1 1.2 Reactor Coolant System Integrity 2.2 1.2-1 SURVEILLANCE LIMITING CONDITION FOR OPERATION REQUIREMENTS 3.1 Reactor Protection System 4.1 3.1-1 3.2 Protective Instrumentation 4.2 3.2-1 A.
Primary Containment Isolation Functions A
3.2-1 8.
Core and Containment Cooling Systems B
3.2-1 C.
Control Rod Block Actuation C
3.2-2 D.
Radiation Monitoring Systems D
3.2-2 E.
Drywell Leak Detection E
3.2-3 F.
Surveillance Information Readouts F
3.2-3 G.
Recirculation Pump Trips and Alternate Rod Insertion G
3.2-4 H.
Accident Monitoring Instrumentation H
3.2-4 3.3 Reactivity Control 4.3 3.3-1 A.
Reactivity Limitations A
3.3-1 l
l B.
3.3-3 C.
Scram Insertion Times C
3.3-6 D.
Reactivity Anomalies 0
3.3-7 E.
Recirculation Pumps E
3.3-7 3.4 Standby Liquid Control System 4.4 3.4-1 A.
Normal System Availability A
3.4-1 B.
Operation with Inoperable Components B
3.4-2 C.
Sodium Pentaborate Solution C
3.4-2 1
3.5 Core and Containment Cooling Systems 4.5 3.5-1 A.
Core Spray and LPCI Subsystems A
3.5-1 B.
Containment Spray Cooling Capability B
3.5-4 i
Amendment No. JJA,151 l
DAEC-1 TECHNICAL. SPECIFICATIONS LIST OF TABLES Table Number Title Page 1.1-1 Deleted 1.1-2 Deleted 1.1-4 Deleted 3.1-1 ReactorProtectionSystem(SCRAM) Instrumentation Requirements 3.1-3 3.1-2 Protective Instrumentation Response Times 3.1-4a 4.1-1 Reactor Protection System (SCRAM) Instrument Functional Tests 3.1-8 4.1-2 Reactor Protection System (SCRAM) Instrument Calibration 3.1-12 3.2-A Instrumentation that Initiates Primary Containment Isolation 3.2-5 3.2-B Instrumentation that Initiates or Controls the Core and Containment Spray System >
3.2-P, 3.2-C Instrumentation that Initiates Control Rod Blocks 3.2-16 3.2-0 Radiation Monitoring Systems that Initiate and/or Isolate Systems 3.2-19 3.2-E Instrumentatior, that Monitors Drywell Leak Detection 3.2-20 l
3.2-F Surveillance Instrumentation 3.2-21 3.2-G Instrumentation that Initiates Recirculation Pump Trip (RPT) and/or Alternate Rod Insertion (ARI) 3.2-23 3.2-H Accident Monitoring Instrumentation 3.2-23a l
4.2.A Minimum Test and Calibration Frequency for PCIS 3.2-24 i
4.2-B Minimum Test and Calibration Frequency for CSCS 3.2-26 4.2-C Minimum Test and Calibration Frequency for Control l
Rod 31ocks Actuation 3.2-28 v
Amendment No. 114. 151
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id DAEC-1 TECHNICAL SPECIFICATIONS LIST OF TABLES (Continued)
Table Number Title Page 4.2-D Minimum Test and Calibration Frequency for Radiation Monitoring Systems 3.2-29 4.2-E Minimum Test Calibration Frequency for Drywell Leak Detection 3.2-30 4.2-F Minimum Test Calibration Frequency for Surveillance Instrumentation 3.2-31 4.2-G Minimum Test and Calibration Frequency for Recircula-tion Pump Trip (RPT) and/or Alternate Rod Insertion (ARI) 3.2-34 4.2-H Accident Monitoring Instrumentation Surveillance Requirements 3.2-34a 3.7-1 Containment Penetrations Subject to Type "B" Test Requirements 3.7-20 3.7-2 Containment Isolation Valves Subject to Type "C" Test Requirements 3.7-22 3.7-3 Primary Containment Power Operated Isolation Valves 3.7-25 4.7-1 Summary Table of New Activated Carbon Physical Properties 3.7-50 4.10-1 Summary Table of New Activated Carbon Physical Properties 3.10-7 3.12-1 Deleted
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3.12-2 Deleted 3.13-1 Fire Detection Instruments 3.13-11 l
l 3.13-2 Required Fire Hose Stations 3.13-12 6.2-1 Minimum Shift Crew Personnel and License Requirements 6.2-3 6.9-1 Deleted l
6.11-1 Reporting Sumary - Routine Reports 6.11-12 6.11-2 Deleted vi Amendment No. JJA,JJ/,151 i
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- DAEC-1 6
- L.
TECHNICAL SPECIFICATIONS LIST OF FIGURES
- Figure.
Number _
Title 11.1 Power / Flow Map 1.1-2 Deleted 2.1-1 APRM Flow Biased Scram and Rod Blocks 2.1-2 Deleted 4.1-1 Instrument Test Interval Determination Curves 4.2-2 Probability of System Unavailability vs. Test Interval 3.4-1 Sodium Pentaborate Solution Volume Concentration Requirements 3.4-2 Minimum Temperature of Sodium Pentaborate Solution l
3.6-1
.DAEC Operating Limits 4.8.C-1 DAEC Emergency Service Water Flow Requirement 3.12-1 Flow-Dependent Minimum Critical Power Ratio (MCPR )
p 3.12 Power-Dependent Minimum Critical Power Ratio Multiplier (K )
p 3.12-3 Minimum Critical Power Ratio (MCPR) versus t (Fuel Types: BP/P8X8R, GE 8X8EB, LTA-311 and ELTA)_
3.12-4 Limiting Average Planar Linear Heat Generation Rate (Fuel Type: BD303A) 3.12-5 Limiting Average Planar Linear Heat Generation Rate (Fuel Type: LTA 311) 3.12-6 Limiting Average Planar Linear > Heat Generation Rate (Fuel Type: BP/PSDRB30lL) t 3.12-7 Limiting Average Planar Linear Heat Generatior. Rate r
(Fuel Type: -BD299A) 3.12-8 Limiting Average Planar Linear Heat Generation Rate (Fuel Types: BP/P80RB299 and ELTA) n 3.12-9 Limiting Average Planar Linear Heat Generation Rate j
(Fuel Tupe: P8DRB284H) i vii Amendment No. J29,U2, 151 i
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DAEC-1 LIMITING CONDITION FOR OPERATIGN SURVEILLANCE REQUIREMENT E.
Drywell Leak Detection E.
Drywell Leak Detection The limiting conditions of Instrumentation shall be operation for the calibrated and checked as instrumentation that monitors indicated in Table 4.2-E.
drywell leak detection are given in Table 3.2-E.
F.
Surveillance Information F.
Surveillance Information Readouts Readouts The limiting conditions for Instrumentation shall be the instrumentation that calibrated and checked as provides surveillance indicated in Table 4.2-F.
information readouts are given in Table 3.2-F.
G.
Recirculatien Pump Trips [_R_PT)
G.
Recirculation Pump Trips and Alternate Rod Insertion and Alternate Rod Insertion (ARI)
(ATWS) - RPT/ARI Instrumentation and logic shall be functionally testcd, The limiting conditions for calibrated, and response time operation for the tested as indicated on Table instrumentation that trips 4.2-G.
the recirculation pumps and initiates Alternate Rod Insertion (ARI) as a means of limiting the consequences of a failure to scram during an anticipated transient are given in Table 3.2-G.
The limiting conditions for operation for the instrumentation that trips the recirculation pumps during turbine stop valve or control valve fast closure for transient margin improvement (especially for end of cycle) are given in Table 3.2-G.
H.
Accident Monitoring H.
Accident Monitoring Instrumentation Instrumentation The limiting conditions for Instrumentation shall be operation for the accident calibrated and checked as monitoring instrumentation are indicated in Table 4.2-H in all given in Table 3.2-H.
operational modes other than COLD SHUTDOWN or refueling.
3.2-4 Amendment No. 197,151
TABLE 3.2-G INSTRUMENTATION THAT INITIATES RECIRCULATION PUMP TRIP (RPT) AND/0R ALTERNATE _R00 INSERTION (ARI) l-Minimum Number of
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Operable Instrument Number of Instruent Channels per Trip Channels Provided System (1)
Instrunent Trip Level Setting By Design Action (ATWS) RPT/ARI 2
Reactor High
< 1140 psig 4*
(2) (6)
Pressure (ATWS) RPT/ARI 2
Reactor low-
> +119.5 in.
4*
(2) (6)
Low Water Tndicated level (5)
Level 1
(E0C) RPT Logic N/A 2
(3) 1 (EOC) (RPT SystemResponse Time)
-< 140 msec (4) 2 (3)
NOTES FOR TABLE 3.2-G 7
1.
Whenever the reactor is in the RUN Mode, there shall be one operable trip system for each parameter.
If g
this cannot be met, the indicated action shall be taken.
2.
With one instrument channel inoperable, rertore the inoperable instrunent channel to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. With both instrunent channels inoperable, restore at least one instrunent channel to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least E
HOT STANDBY within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
E 3.
Two E0C RPT systems exist, either of which will trip both recirculation punps. The systems will be g
individually functionally tested rionthly.
If the test period for one RPT system exceeds two consec.utive hours, the system will be declared inoperable.
If both RPT systems are inoperable or if one RPT system a
is inoperable for more than 72 consecutive hours, an orderly power reduction shall be initiated and tne z
P reactor power shall be less than 85% within four hours.
U 4
This response time is from initiation of Turbine control valve fast closure or Turbine stop valve closure P
to actuation of the breaker secondary (auxiliary) contact.
5.
Zero referenced to top of active fuel.**
6.
If an instrument (s) is inoperable, it may be considered to be OPERABLE if placed in a tripped condition.
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- There are 2 instrunents per trip system that are arranged in a two-cut-of-two once logic.
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- Top of active fuel zone is defined to be 344.5" above vessel zero (see bases 3.2).
TABLE 4.2-G MINIMUM TEST AND CALIBRATION FREQUENCY FOR RECIRCULATION PUMP TRIP (RPT)
AND/0R ALTERNATE R0D INSERTION (ARI)
(ATWS - RPT/ARI)
Instrument Channel
_ Instrument Functional Check Calibration Frequency Reactor High Pressure Annual Annual Reactor Low Water Level Annual Annual System Function Test Frequency Recirculation Pump Trip Once/ operating cycle Alternate Rod Insertion Once/ operating cycle l
(E0C - RPT) l t
g Instrument Channel Functional Check Calibration Frequency Instrument Check Response Time o
{
RPT Initiate Logic Once/ Month N/A N/A N/A O
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RPT Breaker Once/ Operating Cycle H/A N/A Once/ Operating E
Cycle E?
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DAEC-1 timer is set to annunciate before the values specified in Specificatior 3.6.C are exceeded. 'An air sampling system is also provided to detect leakar: inside the primary containment.
For each parameter monitored, as listed in Table 3.2.F. there are two (2) channels of instrumentation. Bycomparingreadingsbetweenthetwo(2) channels, a near continuous surveillance of instrument perfonnance is available. Any deviation in readings will initiate an early recalibra-tion, thereby maintaining the quality of the instrument readings.
On July 26, 1984 the NRC published their final rule on Anticipated Transients Without Scram (ATWS), (10 CFR i 50.62). This rule requires all BWR's to make certain plant modifications to mitigate the consequences of the unlikely occurrence of a failure to scram during an anticipated operational transient. The bases for these modifications are described in NEDE-31096-P-A, "Anticipated Transients Without Scram; Response to NRC ATWS Rule,10 CFR 50.62," December 1985. The Standby Liquid Control System (SLCS) was modified for two-pump operation to provide the minimum required flowrate and boron concentration required by the ATWS rule (seesection3.4 Bases). The existing ATWS Recirculation Pump Trip (RPT) was modified from a one-out-of-two-once logic to trip each recirc. pump to a two-out-of-two-once logic to trip both recire, pumps ("Monticello" design). This logic will also initiate the Alternate Rod Insertion (ARI) system, which actuates solenoid valves that bleed the air off the scram air header, causing the control rods to insert. The instrumer.t setpoints
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are chosen such that the normal reacter protection system (RPS) scram setpoints for reactor high pressure or low water level will be exceeded before the ATWS RPT/ARI setpoints are reached. Because ATWS is considered j
a very low probability event and is outside the normal design basis for j
the DAEC, the surveillance frequencies and LC0 requirements are less stringent than for safety-related instrumentation.
The End-of-Cycle (EOC) recirculation pump trip was added to the plant to improve the operating margin to fuel thermal limits, in particular Minimum Critical Power Ratio (MCPR). The E0C-RPT trips the recire. pumps to lessen the severity of the power increases caused by either a closure of turbine 3.2-45 Amendment No. 151 l
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DAEC-1 stop valves or fast closure of the turbine :ontrol valves with reactor power greater than 30% and a simultaneous failure of the turbine bypass valves to open. The operating limit MCPR of section 3.12.C is calculated assuming an operable E0C-RPT system If the requirements of Table 3.2-G are not mat, then the reactor power level is reduced to a level (86% of rated) which will ensure that the full-power MCPR limits of section 3.12.C will not be violated if such a transient were to occur.
Trip function settings are included for instrument a.c. and battery busses for surveillance of undervoltage relays. The undervoltage relays are required to sense a reduction in the power source voltage so that the subject instruments can be transferred to an alternate power source.
Surveillance tests other than a monthly functional check ot' the bus power monitors for the RHR, Core Spray, ADS, HPCI and RCIC trip systems are not required since they serve as annunciators for complete loss of power and do not monitor reduction of voltage. The subject functional check consists of opening the appropriate circuit breakers and observing the loss of power annunciator activation.
The accident monitoring instrumentation listed in Table 3.2-H were specif-
-ically added to comply with the requirements of NUREG-0737 and Generic Letter 83-36. The instrumentatinn listed is designed to provide plant status for accidents that exceed the design basis accidents discussed in Chapter 15 of the DAEC UFSAR.
Footnote 9 of Table 3.2-H deviates from the guidance of Generic Letter 83-36 l
as continued operation for 30 days (instead of 7 days as recommended in the l
generic letter) is allowed with one of two torus water level monitor (TWLM) channels inoperable. Continued operation is justified by the following l
cond derations:
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Redundancy is available in that at least one channel of the containment water level monitor (CWLM) instrumentation must be available. Since the CWLM envelopes the span measured by the TWLM, the torus water level l
j can be monitored by the CWLM system.
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3.2-45a Amendment No. Jif, 151 l
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...n DAEC-1
.3.4 BASES Standby Liquid Control System 1.
The purpose of the liquid control system is to provide the capa-bility of bringing the reactor from full power to a cold, xenon-free shutdown condition assuming that none of.the withdrawn control rods can be inserted. To meet this objective, the liquid control system is designed to inject a quantity of borun that produces a concentra-tion of 600 ppm of boron in the reactor core in less than 96 minutes based on a minimum 26.2 gpm pump rate. The 600 ppm concentration in l
the reactor core is required to bring the reactor from full power to a subcritical condition, considering the hot to cold reactivity difference, xenon poisoning, analytical biases and uncertainties, etc.
The time requirement for inserting the boron solution was selected to override the rate of reactivity insertion caused by cool-down of the reactor following the xenon poison peak.
The minimum limitation on the relief valve setting is intended to prevent the recycling of liquid control solution via the lifting of a relief valve at too low a pressure. Tne upper limit on the relief valve settings provides system protection from overpressure.
2.
Although the standby liquid control pump circuitry has been modified to run both pumps simultaneously in order to comply with the ATWS 3.4-4 Amendment No. M, 151
1 DAEC-1 rule requirements,* only one of the two standby liquid control pumps is needed for meeting the SLCS design basis. One inoperable pumping circuit does not imediately threaten shutdown capability, and reactor operation can continue while the circuit is being repaired. Assurance that the remaining system will perfonn its intended function and that the long-term average availability of the system is not reduced is obtained for a one-out-of-two system by an allowable equipment out-of-service time of one third of the normal surveillance frequency. This method determines an equipment out-of-service time of ten days. Additional conservatism is intro'duced by reducing the allowable out-of-service time to seven days, and by increased testing of the operable redundant component.
3.
Level indication and alarm indicate whether the solution volume has changed, which might indicate a possible solution concentration range.
The test interval has been established in consideration.
these far. tors. Temperature and 1Iquid level alarms for the system are annunciated in the control room.
l The NRC's final rule on Anticipated Transients Without Scram (ATWS),
10 CFR 5 50.62, requires that the Standby Liq d d Control System (SLCS) be modified to provide an equivalent shutdown capability of 86 gpm at l
13.4 weight percent natural boron for a 251 inch I.D. vessel.
For the DAEC, ATWS equivalence is achieved by running both SLCS pumps simul-taneously at.a minimum combined flow of 45 gpm at a nominal boron concentration of 13% weight percent natural boron, (NEDC-30859, "Duane Arnold ATWS Assessment," December 1984).
(Theequivalenceisalsomet I
if both pumps supply their minimum tech spec flowrate of 26.2 gpm each l
with a solution concentration of at least 11.2 weight percent natural boron.) Because ATWS is a very low probability event and is considered to be beyond the design basis for the DAEC, the surveillance and LCO requirements need not be more stringent than the original SLCS design basis requirements.
3.4-5 Amendment No.151 l
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DAEC-1 The solution is kept at least 5 F above the saturation temperature l
to guard against boron precipitation. The margin is included in Figure 3.4-2.
The volume versus concentration requirement of the solution is such that, should. evaporation occur from any point within the curve, a low level alarm will annunciate before the temperature versus concentration requirements are exceeded.
The quantity of stored boron includes an additional margin (25 percent) beyond the amount needed to shut down the reactor to allow for possible imperfect mixing of the chemical solution in the reactor water.
A minimum quantity of 2500 gallons of solution having a 11.8 percent l
sodium pentaborate concentration, or the equivalent as shown in Figure 3.4-1, will ensure that both the SLCS design basis (11.8%)
and ATWS (11.2%) shutdown requirements are met. The maximum net storage volume of the boron solution, as established by the overflow, is 3270 gallons.
3.4-6 Amendment No. 151
25 r
' ' ' i
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' i Constant Boron Mass Curve 20 (750 ppm Equivalent)
Overflow Vol g
3 Region of Allowable 3:
Vol Conc D
1E g
g 15 g
Ri E t,
.g
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S 0 26.2 gpm Pump Rate
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g m---------------
g =g (11.8%)
y2 z 0>
10 -
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1
, i
, i i r,, f t
i i g
i l
1000 1500 2000 2500 3000
c!
l Net Tank Volume (gal)
DUANE ARNOLD ENERGY CENTER IOWA ELECTRIC LIGHT & POWER COMPANY TECHNICAL SPECIFICATIONS Sodium Pentaborate Solution Volume Concentration Requirements FIGURE 3.4-1 3.4-8 Amendment No. 151
r go Region of Allowable 80 Solution Temperature F
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uinimum soiution
? 70 Temperature Curve 3
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(Sat. Temp. + S*)
3 e
g so
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(5;.C 'F) l l
l l
w l
l l
1 l
I (11.8%)
l l
l l
40 l
10 12 14 18 18 l
Sodium pentaborate Solution (Percent by Weight) l l
I l
DUANE A1.NOLD ENERGY CENTER IOWA ELECTRIC LIGHT & POWER COMPANY TECHNICAL SPECIFICATIONS Minimum Temperature of Sodian Pentaborate Solution l
FIGURE 3.4-2 i
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l 3.4-9 Amendment No.151 i
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