ML20080L942
| ML20080L942 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 02/22/1995 |
| From: | Norrholm L NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20080L947 | List: |
| References | |
| NUDOCS 9503030064 | |
| Download: ML20080L942 (10) | |
Text
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UNITED STATES j
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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. enmaa annt g,*****,,2 IES UTILITIES INC.
CENTRAL IOWA POWER COOPERATIVE CORN BELT POWER C00PERAJ1y1 DOCKET NO. 50-331 DUANE ARNOLD ENERGY CENTER AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 207 License No. DPR-49 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by IES Utilities Inc., et al., dated August 15, 1994, as supplemented on December 21, 1994, and January 20, 1995, 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 i
Commission, C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health i
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 i
E.
The issuance of this amendment is in accordance with 10 CFR Part l
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:
9503030064 950222 PDR ADOCK 05000331 P
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Technical Snecifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 207, 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 isfdb1ce and r
shall'be implemented within 90 days of the dat f issuanc e.
F0 HE NU R
EGULATOR COPMI
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Leif J. Norrholm, Project Director Project Directorate III-3 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of issuance: February 22, 1995 i
ATTACHMENT TO LICENSE AMENDMENT NO. 207 FACILITY OPERATING LICENSE NO. OPR-49 DOCKET NO.-50-331 Replace.the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised areas are indicated by marginal lines.
j Remove Insert 111 iii 3.7-3 3.7-3 3.7-4 3.7-4 3.7-4a t
3.7-12 3.7-12 3.7-29 3.7-29 3.7-30 3.7-30 i
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DAEC-1 o-SURVEILLANCE o
LIMITING CONDITION FOR OPERATION REQUIREMENTS PAGE NO.
3.7 Plant Containment Systems 4.7 3.7-1 A.
Primary Containment Integrity A
3.7-1 B.
Primaq containment Power Operated B
3.7-7 Isolation Valves C.
Drywell Average Air Temperature C
3.7-9 I
D.
Pressure Suppression Chamber - Reactor D
3.7-10 Building vacuum Breakers E.
Drywell - Pressure Suppression Chamber E
3.7-11 Vacuum Breakers F.
Deleted F
3.7-12 G.
Suppression Pool Level and Temperature G
3.7 H.
Containment Atmospheric Dilution H
3.7-15 I.
Oxygen concentration I
3.7-16 J.
3.7-17 K.
Secondary Containment Automatic K
3.7-18 Isolation Dampers L.
Standby Gas Treatment System L
3.7-19 M.
Mechanical Vacuum Pump M
3.7-21 3.8 Auxiliary Electrical Systems 4.8 3.8-1 A.
AC Power Systems A
3.8-1 B.
DC Power Systems B
3.8-3 C.
Onsite Power Distribution Systems C
3.8-5 D.
Auxiliary Electrical Equipment -
D 3.8-5 CORE ALTERATIONS E.
Emergency Service Water System E
3.8-6 3.9 Core Alterations 4.9 3.9-1 A.
Refueling Interlocks A
3.9-1 B.
Core Monitoring B
3.9-5 C.
Spent Fuel Pool Water Level C
3.9-6 D.
Auxiliary Electrical Equipment -
D 3.9-6 CORE ALTERATIONS 3.10 Additional Safety Related Plant 4.10 3.10-1 Capabilities A.
Main Control Room Ventilation A
3.10-1 B.
Remote Shutdown Panels B
3.10-2a 3.11 River Level Jpecification 4.11 3.11-1 AMENDMENT NO. A),444,4A4,4)/,##J,207 111
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4 DAEC-1 i
LIMITING CONDITIONS FOR OPERATION
_,,,,3]QTNEILIANCE REOUIREMENTS 1)
Test Pressure All Type B tests shall be.
performed by local pneumatic j
pressurization of the containment penetrations, either individually or in groups, at a pressure not less than Pa.
.2)
Acceptance Criteria The combined leakage rate of all penetrations subject to Type B and C tests shall be less than 0.60:
La.
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Type C Tests 1)
Type C tests shall be performed on containment isolation valves.
Each valve to be tested shall be closed by normal operation and without any preliminary exercising-or adjustments.
E 2)
Acceptance criteria - The combined' leakage rate for all penetrations subject to Type B and C tests shall be less than 0.60 La.
3)
The leakage from any one main steam isolation valve shall not exceed 100 scf/hr at a test pressure of 24 psig.*
The-i combined maximum pathway leakage rate for all four main steam lines shall not exceed 200 sef/hr at a test pressure of 24 psig.
4)
The leakage rate from any containment isolation valve whose j
seating surface remains water covered post-LOCA, and which is hydrostatically Type C tested, shall be included in the Type C test total.
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l If a main steam isolation valve exceeds 100 sef/hr, it will be restored to s 11.5 scf/hr.
y AMENDMENT NO. //J.797,207 3 7-3 4
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-l r-DhEC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS r
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Periodic Retest Schedule 1)
Type A Test After the preoperational leakage rate tests, a set of three Type A tests shall be performed, at approximately equal intervals 3
during each 10-year service period.
(These. intervals may be extended up to eight months if necessary to coincide with refueling outages.)
The third test of each set shall be conducted when the plant is shut down for the lO-year plant in-service inspections.
The performance of Type A tests U
shall be limited to periods when the plant facility is nonoperational and secured in the shutdown condition under administrative control and in-accordance with the plant safety procedures.
2)
Type B Tests a)
Penetrations and seals of this type (except air locks) shall be leak tested at greater than or equal to 43 psig (P ) during each t
reactor shutdown for major refueling or other convenient interval but in no case at intervals greater than two years.
b)
The personnel airlock shall be
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pressurized to greater than or equal to 43 psig (P ) and leak tested at least once every six (6) months. This test interval may be i
extended to the next refueling l
outage (up to a maximum interval i
between P, tests of 24 months) provided there have been no airlock openings since the last I
successful test at P..
c)
Within three (3) days after securing the airlock when containment integrity is required,
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the airlock gaskets shall be leak tested at a pressure of P,.
P i
i AxEwonEwT nO. Yty,ivi,Wy,207 3.7-4
t DAEC-1 LIMITENG CONDITIONS FOR OPERATION SURVEIt.t u M REOUIRM N TS J.
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Type C Tests Type C tests shall be performed I
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 l
valve leakage integrity testing-r shall be performed at.least once every three months in cedst to.
detect excessive leakage of the purge isolation valve resilient
-l seats.1 The purge system _ isolation valves will be tested in three groups, by penetrations drywell purge exhaust group (CV-4302 and CV-4303), torus purge exhaust group (CV-4300 and CV-4301), and
-l drywell/ torus purge supply group (CV-4307, CV-4308 and CV-4306).
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AnswDurwr mO.I15.JA3,2N.207 3.7-4a
..DAEC.
LIMITING CONDITIGIS FOR OFERATIG6 SURVIILIANCE REOUIREMENTS F.
Deleted F.
Deleted r
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AMENDMENT No. Wi/,if(,207 3.7-12
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DAEC-1 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is allowed to restore the vacuum breaker to OPERABLE status. The 72-hour completion Time takes into account the redundant capability afforded by the remaining breakers, reasonable time for the repairs, and the low probability of an event occurring during this period requiring the vacuum breakers to function.
An open vacuum breaker allows communication between the drywell and suppression chamber airspace, and, as a result, there is the potential for suppression chamber overpressurization due to this bypass leakage if a LOCA were to occur. Therefore, the open vacuum breaker must be closed. The 2-hour completion Time is based on the time required to complete the alternate method of verifying that the vacuum breakers are closed, and the low probability of a DBA occurring during this period.
3.7.F and 4.7.F Bases: Deleted t
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l AMENDMENT NO. A44,207 3.7-29
DAEC-1 e'
3.7.G and 4.7.G BASES Sucoression Pool Level and Temoerature 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 1040 psig. Since all of the gases in the drywell are purged into the pressure suppression chamber air space during a loss-of-coolant 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 suppression chamber and that the drywell volume is purged to the suppression chamber.
Using the minimum or maximum water volumes given in t. i specification, containment pressure during the design basis accident approximately 43 psig which is below the design pressure of 56 psig. The maximum volume of 61,500 8
ft (equivalent to an indicated level of 60%) ensures the clearing loads from SRV discharges are not excessive and do not result in excessive pool swell loads during a Design Bases LOCA.
The minimum volume of 58,900 (equivalent to S
an indicated level of 40%) ft results in a submergence of approximately 3 feet.
Based on Humboldt Bay, Bodega Bay, and Harviken test facility data as utilized in General Electric Company document number NEDE-21885-P and data presented in Nutech document, IES Utilities Inc. document number 7884-M325-002, the following technical assessment results were arrived at:
1.
Condensation effectiveness of the suppression pool can be maintained for both short and long tern phases of the Design Basis AMENDMENT NO.
791 207 3.7-30