ML20096C202
| ML20096C202 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 08/24/1984 |
| From: | Vassallo D Office of Nuclear Reactor Regulation |
| To: | Corn Belt Power Cooperative, Central Iowa Power Cooperative, Iowa Electric Light & Power Co |
| Shared Package | |
| ML20096C203 | List: |
| References | |
| DPR-49-A-105 NUDOCS 8409050154 | |
| Download: ML20096C202 (16) | |
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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. 105 License No. DPR-49 1.
The N'uclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Iowa Electric Light & Power Company, et al, dated April 12, 1984, 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 (1) 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 Specifications 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:
8409050154 840824 PDR ADOCK 05000331 P
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' -(2) -Technical-Specifications The. Technical' Specifications contained in Appendices A and B, as revised through Amendment No.105, 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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.. ;,;.2 Domenic B. Vassallo, Chief Operating Reactors Branch #2 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: August 24, 1984 a
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i ATTACHMENT.T0 LICENSE AMENDMENT N0.105 FACILITY-0PERATING LICENSE NO. DPR-49 D0CKET N0. 50-331 Revise the Appendix "A" Technical Specifications by removing the current pages and inserting the revised pages listed below.
The revised areas are identified by vertical lines.
Remove Insert i
Liv iv vi vi 1.0-8 1.0-8 3.6-6 3.6-6 3.8-5 3.8-5 6.5-3 6.5-3 6.5-7 6.5-7 6.6-1 6.6-1 6.10-1 6.10-1 6.11-3 6.11-3 6.11-4 6.11-4 6.11-5 6.11-5 6.11-6 6.11-6 6.11-7 (deleted) 6.~11-8 (deleted) 6.11-9 (deleted) 6.11-10 (deleted) 6.11-11 (deleted) 6.11-14 (deleted)
PAGE NO.
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 l
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 Environmental Qualification 6.13-1 iv Amendment flo.105 i
}ABLENO.
TITLE PAGE NO.
4.2-D Minimum Test and Calibration Frcquancy for Radiation 3.2-29 i
Monitoring Systems
'4.2-E Minimum Test Calibration Frequency for Drywell Leak 3.2-30 Detection 4.2-F Minimum Test Calibration Frequency for Surveillance 3.2-31 Instrumentation 4.2-G Minimum' Test and Calibration Frequency for 3.2-34 Recirculation Pump Trip 4.2-H Accident Monitoring Instrumentation Surveillance 3.2-34a Requirements 3.6-1 Number of Specimens by Source 3.6-33 4.6-3 Safety Related Snubbers Accessible Ouring Normal 3.6-42 Operation 4.5-4 Safety Related Snubbers Inaccessible During Normal 3.6-44 Operation 4.6-5 Safety Related Snubbers in High Radiation Area During 3.5-48 Shutdown and/or Especially Difficult to Remove 3.7-1 Containment Penetrations Subject to Type "B" Test 3.7-20 Requirements 3.7-2 Containment Isolation Valves Subject to Type "C" Test 3.7-22 Requirements 3.7-3 Primary Containment Power Operated Isolation Valves 3.7-25 4.7-1 Summary Table of New Activated Carbon Physical 3.7-50 Properties 4.10-1 Summary Table of New Activated Carbon Physical 3.10-7 Properties 3.12-1 Deleted 3.12-2 MCPR Limits 3.12-9a i
3.13-1 Fire Detection Instruments 3.13-11 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 Protection Factors for Respirators 6.9-8 6.11-1 Reporting Summary - Routine Reports 6.11-12 6.11-2 Deleted
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t vi Amendment No. 105
1 26.
SURVEILLANCE FREQUENCY Periodic surveillance tests, checks, calibrations and examinations shall be p2rformed within the specified surveillance intervals.
These intervals may be adjusted plus or minus 25%.
The operating 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.
- 28. REACTOR TRIP SYSTEM RESPONSE TIME 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.
- 29. R$ PORTABLE EVENT
. A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50.
r i
1.0-8 Amendment No. 105
c.
' LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT
-2.-
2.
At least.one of the relief.
l valves shall be disassembled a.-
From and after the date-that and inspected each refueling the safety valve function of
^
- outage, one relief valve is made or found to be inoperable, continued reactor operation is
. permissible only during the succeeding thirty -days unless such valve function is sooner made operable; b.
From and after the date that the safety valve function'of two relief valves is made or found to be inoperable, continued reactor operation is L
permissible only during the I
succeeding seven days unless such valve function is sooner made operable.
3.
If Specification 3.6,0.1 is 3.
With the reactor pressure > 100 not met, an orderly shutdown psig and turbine bypass flow to shall be initiated and the the main condenser, each relief-reactor coolant pressure shall valve shall be manually opened be reduced to atmospheric and verified open by turbine within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
bypass valve position decrease and pressure switches and thermocouple readings downstream of the relief valve to indicate steam flow from the valve once per operating cycle.
E.
Jet Pumps E.
Jet Pumos 1.
Whenever the reactor is in the 1.
Whenever there is recirculation startup or run modes, all jet flow with the reactor in the pumps shall be operable.
If startup or run modes, jet pump it is determined that a jet operability shall be checked pump is inoperable, an orderly daily by verifying that the shutdown shall.be initiated following conditions do not and the reactor shall be in a occur simultaneously:
Cold Shutdown Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
a.
The two recirculation loops have a flow imbalance of 15% or more when the pumps are operated at the same speed.
4 Amendment No. 105 3.6-6
5 DELETED Amendment No.105 3.8-5
e.
Investigation of all violations of the Technical' Specifications including the preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence to the Director-Nuclear Generation and to the Chairman
'of the Safety Committee.
f._ Review of all Reportable Events.
l g.
Review of facility operations to detect potential safety hazards.
h.
Performance of special reviews, investigations or analyses and reports thereon as requested by the Chairman of the Safety Committee,
- i. Review of-the Plant Security Plan and implementing procedures.
J.
Review of the Emergency Plan and implementing procedures.
6.5.1.7 Authority The Operations Committee shall:
a.
Recommend to the Plant Superintendent-Nuclear written approval or disapproval of items considered under Specification 6.5.1.6 (a) through (d) above.
6.5-3 Amendment No.105 u.
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6.5.2.7 Review The Safety Committee shall review:
- a. -The safety evaluations for (1) changes to procedures, and (2) tests or experiments ccrupleted under the provision of Section 50.59, 10 CFR, to verify that such actions did not constitute an unreviewed safety-question.
b.
Proposed changes to procedures, equipment or systems which involve an unreviewed safety question as defined in Section 50.59, 10 CFR.
Proposed tests or experiments which involve an unreviewed safety c.
question as defined in Section' 50.59, 10 CFR.
d.
Proposed changes. in Technical Specifications or' licenses.
e.
Violations of applicable statutes, codes, regulations, orders, I
technical specifications, license requirements, or of internal procedures or instructions havin'g nuclear safety significance.
4 f.
Significant operating abnormalities or deviations from normal and i
expected performance of plant equipment that affect nuclear safety.
g.
All Reportable Events.
l h.
All recognized ' indications of an unanticipated deficiency-in some -
aspect of design or operation of safety-related structures, systems, or components.
4
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6.5-7
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l 6.6 REPORTABLE EVENT ACTION l
6.6.1 The following actions shall be taken for REPORTABLE EVENTS.
l Each REPORTABLE EVENT shall be reviewed by the Operations Committee, a.
and a report shall be submitted to the Safety Committee and the Director, Nuclear Generation and b.
The Commission shall be notified and a report submitted pursuant to the requirements of Section 50.73 to 10 CFR Part 50.
6.6-1 Amendment No. 105
6.10 RECORDS RETENTION 6.10.1 The following records shall be retained for at least 5 years:
1.
Records and logs of f acility operation covering time interval at each power level.
2.
Records and logs of principal maintenance activities, inspections, and repair and replacement of principal items of equipment related to nuclear safety.
3.
All Licensee Event Reports.
l 4.
Records of surveillance activities, inspections and calibrati6ns required by these Technical Specifications.
S.
' Records of reactor tests and experiments.
6.
Records of changes made to Operating Procedures.
7.
Records of radioactive shipments.
8.
Records of sealed source leak test and results.
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9.
Records of annual physical inventory verifying accountability of sources on record.
l.
6.10-1 Amendment No. 105
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Monthly Operatino Report Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the U.S. Nuclear Regulatory Commission, l
Washington,-0.C.
20555, 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.
d.
Table 6.11-1 lists some of the routine reports required by 10 CFR Parts 20, 40, 50 and 70, including those listed in Specification 6.11.1.
e.
Annual Safety / Relief Valve Challenge A report documenting safety / relief valve challenges shall be submitted within 60 days i
of January 1 each year.
1 J
Amendment flo.105 6.11-3 g
6.11.2
- Deleted, o
6.11.3.
UNIQUE REPORTING REOUIREMENTS Special reports shall be submitted to the Director of
' Inspection and Enforcement Regional Office within the time period specified for each report.
These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification.
a.
Reactor vessel base, weld and heat affected zone metal test specimens (Specification 4.6.A.2).
b.
I-131 dose equivalent exceeding 50% of equilibrium value (Specification 4.6.8.1.h).
c.
Inservice inspection (Specification 4.6.G).
d.
Reactor Containment Integrated Leakage Rate Test (Specification 4.7. A.2.f).
e.
deleted f.
Fire Protection Systems (Specifications 3.13.A.3, 3.13.8.3, 3.13.C.3, and 3.13.0.3 ).
6.11-4 Amendment No, 105 l
TABLE 6-11-1 REPORTING
SUMMARY
- ROUTINE REPORTS Requirement Report Timing of Submittal TS Annual Safety /
Within 60 days after Relief Valve January 1.
Challenge TS Annual Exposure Within 60 days after January 1.
520.407 Personnel Exposure Within first quarter of and Monitoring each calendar year.
620.408 Personnel Exposure Within 30 days after the on Termination of exposure of the individual Employment or Work has been cetermined or 90 days after date of termination of employinent or work assignment, whichever is earlier.
540.64(a)
Transfer of Source Promptly upon transfer.
Material G40.64(a)
Receipt of Source Within 10 days after Material material is received.
40.64(b)
Source Material Within 30 days after Inventory September 30 of each year.
6.11-5 Amendment No, 105 l
TABLE 6-11-1 (cont)
REPORTING SUMARY - ROUTINE REPORTS Requirement Reoort Timino of Submittal 50.59(b)
Changes, Tests, Within 60 days after and Experiments January 1.
- 70.53 Special Nuclear Within 30 days after March Material Status 31 and September 30 of each year.
570.54 Transfer of Special Promptly up.on transfer Nuclear Material 70.54 Receipt of Special Within 10 days after Nuclear Material material is received Appendix G Fracture Toughness On an individual-case basis to 10 CFR at least 3 years prior to Part 50 the date when the predicted fracture toughness levels will no longer satisfy section V.8. of Appendix G to 10 CFR Part 50.
Appendix H Reactor Vessel Completion of tests after to 10 CFR Material Surveillance each capsule withdrawal.
Part 50 Appendix J Reactor Containment kpproximately3 months to 10 CFR Building Integrated following conduct of test.
Part 50 Leak Rate Test 1Technical Specifications 6.11-6 Amendment No. 105 l