ML20202H232
| ML20202H232 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 07/09/1986 |
| From: | Muller D Office of Nuclear Reactor Regulation |
| To: | Central Iowa Power Cooperative, Corn Belt Power Cooperative, Iowa Electric Light & Power Co |
| Shared Package | |
| ML20202H235 | List: |
| References | |
| DPR-49-A-133 NUDOCS 8607160243 | |
| Download: ML20202H232 (12) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION g
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-t WASHINGTON, D. C. 20555
\\.....,o IOWA ELECTRIC LIGHT AND POWER COMPANY CENTRAL IOWA POWER COOPERATIVE CORN BELT POWER COOPERATIVE DOCKET NO. 50-331 DUANE ARNOLD ENERGY CENTER AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.133 License No. DPR-49 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Iowa Electric Light and Power Company, et al, dated January 9, 1986, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission'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 Commission; 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 Commission'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:
8607160243 860709 DR ADOCK 0500033g PDR
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- 1 (2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.133, 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 j
/W k
Daniel R.'
Muller, Director BWR Project Directorate #2 Division of BWR Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance:
July 9, 1986
+4 84 0 1
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ATTACHMENT TO LICENSE AMENDMENT NO. 133 i
FACILITY OPERATING LICENSE NO. DPR-49 DOCKET NO. 50-331 Replacethefoll$wingpagesoftheAppendixATechnicalSpecificationswith the enclosed pages.
The revised areas are indicated by marginal lines.
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Pages iv j
1.1-4 3.2-8 3.5-17 3.7-13 i
3.8-10 l
6.10-3 6.11-15 (deleted) j 6.13-1 (deleted) 4 4
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DAEC-1 PAGE NO.
I 5.0 Design Features 5.1-1 5.1 Site 5.1-1 5.2 Reactor 5.2-1 5.3 Reactor Vessel 5.3-1 5.4 Containment 5.4-1 5.5 Spent and New Fuel Storage 5.5-1 6.0 Administrative Controls 6.1-1 6.1 Management - Authority and 6.1-1 Responsibility 6.2 Plant Staff Organization 6.2-1 6.3 Plant Staff Qualifications 6.3-1 6.4 Retraining and Replacement Training 6.4-1 6.5 Review and Audit 6.5-1 6.6 Reportable Event 6.6-1 6.7 Action to be Taken if a Safety Limit 6.7-1 is Exceeded 6.8 Plant Operating Procedures 6.8-1 6.9 Radiological Procedures 6.9-1 6.10 Records Retention 6.10-1 6.11 Plant Reporting Requirements 6.11-1 6.12 Deleted 6.13 Deleted l
I Amendment No. k, 133 gy
DAEC-1 l
SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING l
E.
Scram - main
< 10 percent steam line valve closure isolation valve F.
> 850 psig isolation valve closure nuclear system low pressure G.
> 363 inches
& LPCI ibove vessel actuation -
zero (+18.5 reactor low inches water level indicated level) 1 H.
> 464 inches actuation -
ibove vessel reactor low zero(+119.5 water level inches indicated
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1evel)
J I.
> 363 inches 4.
isolation above vessel valve closure-zero (+18.5 i
reactor low inches
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water level indicated level.)
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< 10 inches Hg isolation vacuum valve closure-loss of main condenser vacuum t
e Amendment No.[ % 133 1.1-4
1 TABLE 3.2-B g(
INSTRUMENTATION THAT INITIATES OR CONTROLS THE CORE AND CONTAINMENT COOLING SYSTEMS 3
Minimum No.
of Operable g
Instrument Number of Channels Per Instrument Channels Trip System (1)
Trip Function Trip Level Setting Provided by Design Remarks 2
Reactor Low-Low 1 + 119.5 in. indicated 4 HPCI & RCIC Initiates HPCI & RCIC Water Level level (4)
Instrument Channels Initiates LPCI loop select logic tj 4 LPCI loop select Instrument Channels 2
Reactor Low-Low-Low 2. + 18.5 in. indicated 4 Core Spray & RHR
- 1. In conjunction with Water Level level (4)
Instrument Channels Low Reactor Pressure initiates operation 4 ADS Instrument of Core Spray and Channels LPCI valves. Starts f[
pumps if not already g,
started from 2 psig dryweil signal.
- 2. In conjunction with confinnatory low level, 120 second l
time delay and LPCI or Core Spray pump interlock initiates Auto Blowdown (ADS)
- 3. Initiates starting of Diesel Generator
- 4. Closes group 7 isolation valves 2
Reactor High Water j[ + 211 in. indicated 2 Instrument Channels Trips HPCI and RCIC Level level (4) turbines
DAEC-1 1 LPCI pump must be available to fulfill the containment spray function. The 7 day repair period is set on this basis.
B&C Containment Spray and RHR Service Water The containment spray subsysten for DAEC consists of 2 loops each with 2 LPCI pumps and 2 RHR service water pumps per loop.
The, design of these systems is predicted upon use of 1 LPCI, and 2 RHR service water pumps for heat removal after a design basis event. Thus, there are ample spares for margin above the design conditions.
Loss of margin should be avoided and the equipment maintained in a state of operability so a 30-day out-of-service time is chosen for this equipment.
If one loop
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is out-of-service, or one pump in each loop is out-of-service, reactor operation is permitted for seven days with daily testing of the operable loop (s) af ter testing the appropriate diesel generator (s).
With components or subsystems out-of-service, overall core and containment cooling reliability is maintaired by demonstrating the operability of the remaining cooling equipment.
The l
degree of operability to be demonstrated depends on the l
nature of the reason for the out-of-service equipment. For routine out-of-service periods caused by preventative Amendment No. M i33 3.5-17 i
'5IMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 6.
Containment Atmosphere 6.
Containment Atmospherc
' Dilution 511ution a.
Whenever the reactor is -
a.
The post-LOCA containment i
in power operation, the atmosphere dilution system Post-LOCA Containment shall be functionally tested Atmosphere Dilution System once per operating cycle, must be operable and capabic of supplying nitro-gen to the containment for atmosphere dilution if required by post-LOCA conditions.
If this spec-ification cannot be met, thc. system must be restored to an operable condition within 7 d.ays or, the reactor must be taken out of power opdration.
The volume in the N b.
Whenever ~the reactor is in b.
power operation, the post-storagebankshallbe LOCA Containment Atmos'phere recorded weekly.
i Dilution System shall con-
- ' tain a minimum of 50,000 scf of N2 as determined by j
pressure and temperaturc measurements.- If this specification cannot be a,
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met, the minimum volume 4
will be restored within.7 days or the reactor must,
i be taken out of power operation, except that for 1
the period fron February 25, 1975 l
to February 26, 1975, the mini--
mum volune will be restored within 9 days or the reactor
- must be taken out of power operation.
Whenever the reactor is in power c.
The CAD system H2 and 02 c.
t operation, there shall be at analyzers shall be tested 1 cast one CAD system Il 'and 0 3
2 for operability using analyzer serving the dry.fc11 standard bottled H2 and 02 and the suppression chamber.
If once per month and shall be this specification cannot be met calibrated once per 6 the reactor must be ta, ken out of,
months.
The atmosphere Power operation.
analyzing system shall be I
j l3.7-13 r
i Amendment No. S 133
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.P
DAEC-1 The 250 volt d-c system provides power for the HPCI system.
If the battery is taken out of service, the HPCI system would be inoperable l
and the requirements of Specification 3.5.D for this condition must be satisfied.
The 24 volt d-c system provides power for source range monitoring, intermediate range monitoring, and liquid process radiation monitoring, i
The two neutron monitoring functions are required for safety, however, the design is fail-safe in that loss of 24 volt d-c power would cause the associated trip to function (UFSAR Section 8.3.2).
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a The battery roam is ventilated to prevent accumulation of hydrogen gas exceeding 4 percent concentration. On loss of battery room
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ventilation, the use of portable ventilation equipment and daily
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sampling provides assurance that potentially hazardous quantities of hydrogen gas will not accumulate.
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Amendment No. 1,14', 133 3.8-10 '
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DAEC-1 l
7.
Records of training and qualification for current members of the plant staff.
1 8.
Records of in-service inspections performed pursuant to these Technical i
Specifications.
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9.
Records of Quality Assurance activities required by the QA Manual with the l
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exception of the records included in Section 6.10.1.
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Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59.
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Records of meetings of the Operations Committee and the Safety Committee, i
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1 12.
Records of the service lives of all safety-related hydraulic and mechanical l
snubbers including the date at which the service life commences and i
associated insta'ilation and maintenance records, i
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13.
Records of results of analyses required by the radiological environmental l
I monitoring program.
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I Amendment No. M IM M 133 6.10-3
DAEC-1 i
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THIS PAGE DELETED i
Amendment No.133 6.11-15 I
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DAEC-1 NIS PAGE DELETED Amendment No. 133 6.13-1
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