ML20087P118
| ML20087P118 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 03/08/1984 |
| From: | Vassallo D Office of Nuclear Reactor Regulation |
| To: | Iowa Electric Light & Power Co, Central Iowa Power Cooperative, Corn Belt Power Cooperative |
| Shared Package | |
| ML20087P119 | List: |
| References | |
| DPR-49-A-094 NUDOCS 8404060072 | |
| Download: ML20087P118 (8) | |
Text
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IOWA ELECTRIC LIGHT AND POWER COMPANY CENTRAL IOWA F0WER COOPERATIVE CORN BELT POWER COOPERATIVE DOCKET N0. 50-331 DUANE ARNOLD ENERGY CENTER AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 94 License No. DPR-49 4
1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Iowa Electric Light & Power Company, et al, dated Novenber 30, 1976, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; 8.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulatinns of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can he 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; D.
The issuance of this amendment will not be inimical to the common 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 Commission'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 follows:
8404060072 840300 PDR ADOCK 05000331 P
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 94, 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.
FOR THE NUCLEAR REGULATORY COMMISSION
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Domenic B. Vassallo, Chief Operating Reactors Branch #2 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance:
March 8, 1984 l
1
ATTACHMENT TO LICENSE AMENDMENT NO. 94 FACILITY OPERATING LICENSE NO. DPR-49 DOCKET N0. 50-331 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contains vertical lines indicating the area of changes.
AFFECTED PAGES v
1.0-8 3.1-1 3.1-4a 3.1-16 l
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DAEC-1 TECHNICAL SPECIFICATIONS LIST OF TABLES TABLE NO.
TITLE PAGE N0.
1.1-1 Deleted 1.1-2 Deleted 1.1-4 Deleted 3.1-1 Reactor Protection System (SCRAM) Instrumentation 3.1-3 Requirements 3.1-2 Protective Instrumentation Resoonse Times 3.1-4a l
4.1-1 Reactor Protection System (SCRAM) Instrument 3.1-8 Functional Tests 4.1-2 Reactor Protection System (SCRAM) Instrument 3.1-12 Calibration 3.2-A Instrumentation that Initiates Primary Containment 3.2-5 Isolation 3.2-B Instrumentation that Initiates or Controls the 3.2-8 Core and Containment Spray Systems 3.2-C Instrumentation that Initiates Control Rod Blocks 3.2-16 3.2-0 Radiation Monitoring Systems that Initiate and/or 3.2-19 Isolate Systems 3.2-E Instrumentation that Monitors Drvwell Leak Detection 3.2-20 3.2-F Surveillance Instrumentation 3.2-21 3.2-G Instrumentation that Initiates Recirculation Pump Trip 3.2-23 3.2-H Accident Monitoring Instrumentation 3.2-23a 4.2-A Minimum Test and Calibration Frequency for PCIS 3.2-24 4.2-8 Minimum Test and Calibration Frequency for CSCS 3.2-26 4.2-C Minimum Test and Calibration Frequency for Control 3.2-38 Rod Blocks Actuation l
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DAEC-1 26.
SURVEILLANCE FRE0VENCY Periodic surveillance tests, checks, calibrations and examinations shall be performed within the specified surveillance intervals.
These intervals may be adjusted plus or minus 25%.
The ooerotinn cycle interval as pertaining to instrument and electrical surveillance shall never exceed 15 months.
In cases where the elapsed interval has exceeded 100% of the specified interval, the next surveillance interval shall commence at the end of the original specified interval.
27.
FIRE SUPPRESSION WATER SYSTEM A fire suppression water system shall consist of a water source, pumps, and distribution piping with associated sectionalizing control or isolation valves.
Such valves include yard hydrant curb valves, the first valve ahead of the water flow alarm device on each sprinkler, hose standpipe or deluge system riser.
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- 28. REACTOR TRIP SYSTEM RESPONSE TIME i
Reactor trip system response time is the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until deenergization of the scram pilot valve solenoids.
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1.0-8 AMENDMENT NO. 94
DAEC-1
.s LIMITING CONDITIONS FCR OPERATION SURVEILLANCE RE0VIREMENT 3.1 REACTOR PROTECTION SYSTEM 4.1 REACTOR PROTECTION SYSTEM Applicability:
Apolicability:
Aoplies to the Applies to the surveillance of the instrumentation and instrumentation and associated associated devices which devices which initiate reactor initiate a reactor scram.
Ob.iective:
Ob.iective:
To assure the operability of To specify the type and frequency the reactor protection system.
of surveillance to be applied to the protection instrumentation.
Specification:
Specification:
A.
The setpoints, minimum number A.1 Instrumentation systems shall be of trip systems, and minimum functionally tested and calibrated number of instrument channels as indicated in Tables 4.1-1 and that must be operable for 4.1-2 respectively.
each position of the reactor mode switch shall be as given
.2 Response time measurements (from in Table 3.1-1.
The designed actuation of sensor contacts or system response times from trip point to de-energization of the opening of the sensor scram solenoid relay) are not part contact uo to and including of the normal instrument the openina of the trio calibration. The reactor trip actuator contacts shall not system response time of each exceed 50 milliseconds.
reactor trip function shall be demonstrated to be within its As a minimum, the reactor limit at least once oer 18 months.
protection system Each test shall include at least instrumentation channels of one logic train such that both Table 3.1-1 shall be operable loaic trains are tested at least with response times as shown once per.36 months and one channel in Table 3.1-2.
per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip function.
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.3 Daily during reactor power operation, the MFLPD and the FRP shall be checked and the APRM SCRAM and APRM Rod Block settings given by equations in Specification 2.1.A.1 and 2.1.B shall be calculated if the MFLPD exceeds the FRP.
.4 When it is determined that a channel has failed in the unsafe condition, the other RPS channels that monitor the same' variable shall be functionally AMENDMENT N0. 94
DAEC-1 9
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TABLE 3.1-2 PROTECTIVE INSTRUMENTATION RESPONSE TIMES l
Reactor Sensor Trip System j.
Functional Unit Response Time Response Time i
l 1.
Reactor Vessel Steam Dome Pressure - High
<.5 seconds 1
55 seconds 2.
Reactor Vessel Water Level - Low
<1.0 seconds i 1.05 seconds 1
1 4
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3.1-4a AMENDMENT NO. 9a i
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DAEC-1 from the manual scram push buttons and the reactor mode switch.
Each remainino subchannel has an input from at least one independent instrument channel which monitors a critical oarameter.
The outputs of the subchannels are combined in a 1 out of 2 looic:
i.e.,
an input signal on either one or both of the subchanrals will cause a trip system trip.
The outputs of the trip systems are arranged so that a trio on both trip systems is required to produce a reactor scram.
This system meets the intent of IEEE - 279 for Nuclear Power Plant Protection Systems.
The system has a reliability greater than that of a 2 out of 3 system and somewhat less than that of a 1 out of 2 system.
The measurement of response time at the specified frequencies provides assurance that the protective, isolation and emergency core cooling functions associated with each channel is completed within the time limit assumed in the accident analysis.
Response time may be demonstrated by any series of sequential, overlapping or total channel test measurements, provided such tests demonstrate the total channel response time as defined.
Sensor response time verification may be demonstrated by either: 1) inplace on-site or off-site test measurements, or 2) utilizing replacement sensors with certified response times.
With the exception of the Average Power Range Monitor (APRM) channels, the-Intermediate Range Monitor (IRM) channels, the Main Steam Isolation Valve closure and the Turbine Stoo Valve closure, each subchannel has one instrument channel.
When the minimum condition for operation on the number of operable instrument channels per untrinned orotection trip system is met or if it cannot be met and the affected protection trip system is placed in a tripped condition, the effectiveness of the protection system is preserved.
3.1-16 AMENDMENT No. 94
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