ML20087P058
| ML20087P058 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 03/13/1984 |
| From: | Vassallo D Office of Nuclear Reactor Regulation |
| To: | Iowa Electric Light & Power Co, Central Iowa Power Cooperative, Corn Belt Power Cooperative |
| Shared Package | |
| ML20087P059 | List: |
| References | |
| DPR-49-A-095 NUDOCS 8404050502 | |
| Download: ML20087P058 (8) | |
Text
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panoog'o UNITED STATES
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NUCLEAR REGULATORY COMMISSION
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'%f. ; p 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. 95 License No. DPR-49 1.
The Nuclear Regulatory Comission (the Commission) has found that:
A.
The application for amendment by Iowa Electric Light & Power Company, et al, dated December 20, 1982, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regula-tions 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 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 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. OPR-49 is hereby amended to read as follows:
8404050502 840313 PDR ADOCK 05000331 P
PDR m
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. (2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.
95, 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 4
Domenic B. Vassallo, Chief Operating Reactors Branch #2 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance:
March 13,1984 L
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ATTACHMENT TO LICENSE AMENDMENT NO. 95 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 pages are identified by Amendment number and contain vertical lines indicating the area of changes.
AFFECTED PAGES 3.7-31 3.7-32 3.7-32a 6.5-3 6.5-5 6.5-6 6.5-8
The pressure suppression pool water provides the heat sink for the reactor primary system energy release following a postulated rupture of the system.
The pressure suppression chamber water volume must absorb the associated decay and structural sensible heat released during primary system blowdown from 1000 psto.
Since all of the gases in the drywell are Durced into the pressure suppression chamber air space durinq a loss-of-colant accident, the pressure resulting from isothermal compression plus the vapor pressure of the liquid must not exceed 62 psig, the suppression chamber maximum allowable pressure.
The design volume of the suppression chamber (water and air) was obtained by considering that the total volume of reactor coolant to be condensed is discharged to the suppres'sion chamber and that the drywell volume is purged to the~ suppression chamber.
Using the minimum or maximum water volumes given in the-specification, containment pressure during the design basis accident is approximately S4 psig which is below the design pressure of 56 psig.
The minimum volume of results in a submergence of approximat"ely 3 feet.
Based on 3
58,900 ft Humboldt Bay, Bodega Bay, and Marviken test facility data as utilized in General Electric C.ompany document number NE0E-21885-P and data presented in Nutech documeni., 'owa Electric document number 7884 M325-002, the following technical assessment results were arrived at:
1.
Condensation ef fectiveness of the suppression pool can be maintained for both short and long term phases of the Design Basis Accident (08A), Intermediate Break Accident (18A), and Small Break j
l Accident (SBA) cases with.three feet submergence.
l
.i 3.7-31 Amendment No. 95
A DAEC-1 2.
There is no significant thermal stratification in the condensation oscillation regime after LOCA with three feet submercence.
3.
There is some thermal stratification in the chuqainq reatme for all break sizes.
However, this will not inhibit the pressure suppression function of the suporession pool.
4.
Seismic induced waves will not cause downcomer vent uncovering with three feet submergence.
5.
Post-LOCA pool waves will not cause downcomer vent uncovering with three feet submergence.
6.
Maximum post-l.0CA drawdown will not cause downcomer vent uncovering and condensation effectiveness of the suppression pool will be maintained.
Therefore, with respect to dnwncomer submergence, this specification is adequate.
The max,imum temperature at the end of blowdown tested during the Humbolt Bay and Bodega Bay tests was 170*F and this is conservatively taken to be the limit for complete condensation of. the reactor coolant, although condensation would occur for temperatures above 170*F.
Should 'it be necessary to drain the suppression chamber, this should only be done when there is no requirement for core standby. cooling systems operability as explained in. Basis 3.5.G or the requirements of 9pecification 3.5.G.4 are met.
Amendment No. 95
i a
DAEC-1 Using a 50*F rise (Table 5.2-1, FSAR) in the suppression chamber water temperature and a minimum water volume of 58,900 ft, the 170*F temperature 3
which is used for complete condensation would be approached only if the suppression pool temperature is 120'F prior to the DBA-LOCA.
Maintafning a pool temperature of 95'F will ~ssure that the 170'F limit is not approached.
a 2.
Inerting Safety Guide No. 7 essumptions for metal-water reactions result in hydrogen concentrations in excess of the Safety l
l 3.7-328 Amendment No. 95 l
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DAEC-1 Investigation of all violations of the Technical Specifications including the e.
preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence to the Director-Huclear Generation and to the Chairman of the Safety Committee.
f.
Review of those Reportable Occurrences requ' iring 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> notification to the Commission, 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
i.
Review of the Pl. ant Security Plan and implementing procedures.
,i, Review of the Emergency Plan and implementing procedures, e
6.5.1.7 Authority The Operations Committee shall:
Recommend to the Plant Superintendent-Nuclear written aporoval or disapproval a.
ofitemsconsideredunderSpecification6.5.1.6(a)through(d)above.
6.5-3 Amendment No. 95
DAEC-1
- c.
Chemistry and radiochemistry.
d.
lietallurgy.
\\
1 e.
Instrumentation and control.
f.
Radiological safety.
g.
liechanical and electrical engineering, j
h.
Quality assurance practices, i.
Non-destructive testing.
,i.
Administration.
6.5.2.2 Comcosition i'
The Safety Committee shall be composed of persons who have been appointed in writing by the President to serve on a permanent basis and who collectively have or have access to applicable technica) and experimental expertise in the areas listed in section 6.5.2.1, items a through j.
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^6.5-5 Amendmer.t No. 95-
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O DAEC-1 l
6.5.2.3 Alternates All alternate members shall tvi appointed in writinq by the President to serve on a permanent basis.
6.5.2.4. Consultants Consultants shall be utilized as determined by the Safety Committee Chairman to provide expert advice to the Safety Committee.
6.5.2.5 Heeting Frequency The Safety Committee shall meet at least once per calendar quarter during the initial year of facility operation following fuel loading and at least once per six months thereafter, e
6.5.2.6 Ouorum A quorum of the Safety Committee shall consist of the Chairman or Vice Chairman and at least four members with a maximum of two alternates as votino members.
No more than a minority of the voting members shall have line responsibility for operation of the facility.
6.5-6 Amendment No. 95 I
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DAEC-1 1.
Reports and meeting minutes of the Operations Committee.
6.5.2.8 Audits Audits of facility activities shall be performed under the cognizance of the Safety Committee. These audits shall encompass:
The conformance of facility operation to all provisions contained a.
within the Technical Specifications and applicable license conditions at least once per 24 months.
The performance, training and qualifications of the entire facility b.
staff at least once per 24 months, The results of all actions taken to correct deficiencies occuring in c.
facility equipment, structures, systems or method of operation that affect nuclear safety at least once per six months, d.
The performance of all activities required by the Quality Assurance Program to meet the criteria of Appendix "8",10CFR50, at least once per 24 months.
I The Emergency Plan and implementing procedures at least once per 12 e.
months.
f.
The Security Plan and implementing procedures at least once per 12 months.
6.5-8 Amendment No.- 95