ML20091J713
| ML20091J713 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 05/22/1984 |
| From: | Vassallo D Office of Nuclear Reactor Regulation |
| To: | Corn Belt Power Cooperative, Central Iowa Power Cooperative, Iowa Electric Light & Power Co |
| Shared Package | |
| ML20091J716 | List: |
| References | |
| DPR-49-A-100 NUDOCS 8406060121 | |
| Download: ML20091J713 (9) | |
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WASHINGTON, D C. 20555 y.....J 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.100 License No. DPR-49 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Iowa Electric Light & Pcwer Company, et al, dated June 10, 1982, 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; 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 (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 Comission's regulations; 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 Speciff-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:
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. (2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.100, 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 its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION f?f4s)L Domenic B. Vassallo, Chief Operating Reactors Branch #2 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: May 22,1984 Y
a
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ATTACHMENT TO LICENSE AMENDMENT NO.100 FACILITY OPERATING LICENSE NO. DPR-49 DOCKET NO. 50-331 Revise the Appendix "A" Technical Specifications as noted below:
Remove Insert 3.7-7 3.7-7 3.7-7a 3.7-14 3.7-14 3.7-14a 3.7-14a 6.11-11 6.11-11 The revised pages are identified by Amendment number and contain vertical o
lines indicating the areas of change.
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DAEC-1 LIMITINGCONDITIONSFORONERATION SURVEILLANCE REOUIREMENT 3)
Tyoe C Tests Type C tests shall be performed during each reactor shutdown for major refueling or other convenient interval but in no case at intervals greater than two years.
4)
Additional Periodic Tests Additional purge system isolation valve leakage integrity testing shall be performed at least once every three months in order to detect excessive leakage of the purge isolation valve resilient seats.
The purge system isolation valves will be tested in three groups, by penetration: drywell purge exhaust group (CV-4302 and CV-4303), torus purge exhaust group (CV-4300 and CV-4301), and drywell/ torus purge supply group (CV-4307, CV-4308 and CV-4306).
e.
Seal Reolacement The T-ring inflatable s"eals for purge isolation valves CV-4300, CV-4301, CV-4302, CV-4303, CV-4306, CV-4307 and CV-4308 shall be.
replaced at intervals not to exceed four years.
The baseline for this requirment shall be ' established during the 1982 refueling outage.
f.
Containment Modification Any major modification, replacement of a component which is part of the primary reactor containment boundary, or resealing a seal-welded door, performed after the
. precoerational leakage rate test snall-be followed by eitner a Type A, Tyoe 3, or Type C test, as.
a:alicanie, f:r the area affected
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DAEC-1 o.
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENT The measured leakage from this test shall be included in the test report.
The acceptance criteria as appropriate, shall be met.
Minor modifications, replacements, or resealing.of seal welded doors, performed directly prior to the conduct of a scheduled Type A test oo not require a separate test.
g.
Reporting l
The preoperational and periodic tests shall be the subject of a summary. technical report submitted to the Commission ^
approximately three months after the conduct of each test.
The report will be titled " Reactor Containment Integrated Leakage Rate Test."
, The results of the periodic testing performed'to satisfy the requirements of 4.7.A.2.d.4 shall be reported fith,the
- summary technical report prepared to provide the results of the testing performed in accordance with.Section 4.7.A.2.d.3.
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_3.7-7a Amendment No. 100
DAEC-1
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LIMfTING _CONDIT20NS FOR OPERATf 0N SURVEILLANCE REQUIREMENT must be taken out of power functionally tested once per operation operating cycle in conjunction with specification 4.7. A.S.a.
Should one of tne two H2 or 02 analyzers serving the crywell or
' suppression pool be found inoperable, the remaining analyzer of the same type serving the same compartment shall be tested for operability once per week until the defective analyzer is made operable.
7.
Drywell-Suppression Chamber 7.
Drywell-Suppression Chamber Differential Pressure Offferential pressure a.
Differential pressure between the a.
The pressure differential between drywell and suppression chamber the drywell and suppression shall be maintained at equal to or-chamber s'nall be recorded at least greater than 1.10 psid except as once each shift.
specified-in (1) and (2) below:
(1) Within the 24-hour period subsequent to placing the reactor in the Run Mode following a shutdown, the differential shall be established.
The differential may be decreased to less than 1.10 psid 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior a scheduled shutdo'wn.
(2) This differential may be decreased to less than 1.10 psid for a maximtsn of four hours during required operability testing of the HPCI system pump, the RCIC system pump, the drywell-pressure suppression chamber vacuum breakers, the suppression chamber to reactor l
butiding vacutra breakers, and to perform leak rate testing required by specification 4.7.A.2.d.4, and to allow for inerting operations to satisfy specification 3.7.A.5 requirements.
be If the differential pressure of i
specification 3.7. A.7.a cannot be maintained, and the differential pressure.cannot be restored within i
the subsequent six (6) hour period, an orderly shutdown sna11 be initiated and the reactor snall b in the Cold Shutcown conditica i-witnin tne followin; 2' nour:.
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3.7 14 Amendment No. 100
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DAEC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENT 8.
If the specifications of 3.7.A.1 through 3.7. A.5 cannot be met, an orderly shutdown shall be initiated and the reactor shall be in a cold snutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
I 9.
Durging The time which containment vent / purge valves (CV-4302, CV-4303, CV-4300, CV-4301 and CV-4307) can be open is limited to a maximun of 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> per calendar year, not including the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period prior to : shutdown and the 24
. hour period subsequent to.
placing.the reactor in the y run mode following a shutdown as specified in 3.7. A.S.b.
This restriction applies
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whenever primary containment integrity is required.
10.
If Specification 3.7. A.9 cannot -be met, prepare and submit a Special Report to the Commission ' pursuant to Specification 6.11.3 within the next 30 days outlining the cause of the limits being exceeded and the plans for limiting the time which these valves.will be open, e
't I-3.7-14a Amendment No.
100
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Reactor vessel bas.e, weld and heat affected zone metal
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a.
test specimens (Specification 4.6.A.2).
t b.
1-131 dose equivalent exceeding 50% of equilibri$n value (Specification 4.6.B.l.h).
Inservice inspection (Specification 4.6.G).
.c.
d.
Reactor Containment Integrated ' Leakage Rate Test (Specification 4.7. A.2.f).
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Auxiliary Electrical System - Operation with inoperable e.
components (Specification 3.B.B.4).
L f.
Fire Protection Systems-(Specifications 3.13.A.3, 3.13.B.2, 3.13.B.3, 3.13.C.3, and 3.13.D.3).
g.
Containment Vent / Purge valves (Specification 3.7. A.10).
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T 6.11-11 AmendmentNo.f[,100,
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