ML20108C772
| ML20108C772 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  (DPR-049) |
| Issue date: | 10/29/1984 |
| From: | Vassallo D Office of Nuclear Reactor Regulation |
| To: | Corn Belt Power Cooperative, Central Iowa Power Cooperative, Iowa Electric Light & Power Co |
| Shared Package | |
| ML20108C774 | List: |
| References | |
| DPR-49-A-108 NUDOCS 8411190066 | |
| Download: ML20108C772 (14) | |
Text
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UNITED STATES if.
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. NUCLEAR REGULATORY COMMISSION WASHINGTON. D. C. 20555
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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. 108 License No. DPR-49 1.
The Nuclear Regulatory Comission (the Comission) has found that:
1 A.
The application for amendment by Iowa Electric Light & Power Company, et al, dated July 20, 1983, as supplemented January 27, 1984 and August 8, 1984, 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 (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 Comission'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 Comission'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:
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8411190066 841029 PDR ADOCK 05000331 p
PDR i.
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. (2) Technical Specifications The Technical Specifications contained in Appendices A and B,
-as revised through Amendment No. 108, are hereby incorporated in the license. The licensee shall operate the fa'cility in accordance with the Technical Specifications.
3.
The license anendment-is effective as of the date of issuance.
FOR THE NUCLEAR REGilLATORf COMMISSI0li
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Domenic B. Vassallo, Chief Operating Reactors Branch #2 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: October 29, 1984 S
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ATTACHMENT TO LICENSE AMENDMENT NO.108 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 areas of change..
AFFECTED PAGES 3.5-4 3.5-5 3.5-17 3.5-18 3.5-26 3.7-1 3.7-2 3.7-32 3.7-32a 3.7-32b 3.7-49 I
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d DAEC-l' LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 6.
If the recuirements of 3.5. A 6.
- Once per shift visually inspect -
cannot be met, an orderly and verify that RHR valve panel shutdown of the reactor shall lights and instrumentation are be initiated.and the reactor functioning normally.
shall be in the Cold Shutdown Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
B.
Containment Soray Cooling 8.
Containment Soray Cooling Laoaoiitty Laoaoisity Surveillance of the drywell spray loops shall be performed as follows:
1.
Containment cooling spray-1.
During each five year period,d loops are required to be an air test shall be performe operable when the reactor on the drywell and suppression water temperature is greater pool spray headers and nozzles.
a than 212*F except that a maximum of one drywell spray loop may be inoperable for thirty days when the reactor water temperature is greater than 212*F.
2.
If this requirement cannot be met, an orderly shutdown shall be initiated and the reactor shall be in the Cold Shutdown Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
C.
C.
Surveillance of the RHR Service dervice water system water system 1.
Except as specified in 1.
Surveillance of the RHR service 3
3.5.C.2, 3.5.C.3, 3.5.C.4, water system shall be as 3.5.C.5 and 3.5.G.3 below, folldws:
bothRHkservicewater subsystem loops shall be RHR Service Water Subsystem operable whenever irradiated Testing:
i fuel is in the reactor vessel t
and reactor coolant Item Frecuency temperature is greater than 212*F.
a)
Pump and motor Once/3 months operated valve i
operability.
b)
Flow Rate after major Test-Each pump RHR service maintenance water pump and every 3 shall deliver months at least 2040 i
gpm at a TDH of 610 ft. or more.
3.5-4 Amendment No. 108
DAEC-1 LIMITING CONDITION FOR OPERATION' SURVEILLANCE REOUIREMENT' 2
From and after the date that 2.
Whert it is determined that one one of the RHR Service Water-RHR Service Water pump is subsystem pumps is made or inoperable, the remaining found to be inoperable ~for any components of that. subsystem i
reason, reactor operation must and the other subsystems 'shall be limited to thirty days be demonstrated to be operable unless operability of that.
imediately and daily pump is restored within this thereafter.
period.
During such thirty days all other active components of the RHR Service Water subsystem are operable.
3 3.
From and after the date that I.
When one RHR Service Water pump one RHR Service Water pump in in each subsystem becomes each subsystem is made or' inoperable, the remaining found to be inoperable for any components of both subsystems reason, reactor operation is and their associated diesel-limited to seven days unless generators required for operability of at least one operation of such components, pump is restored within this shall be demonstrated.to be period.
During such seven operable immediately. The days all active components of remaining components of both both RHR Service Water subsystems shall be demonstrated to be operable subsystems and their daily thereafter.
4 associated diesel generators required for operation of such components (if no external source of power were j
available), shall be operable.
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4.
From and after the date that 4.
When one RHR Service Uater l
l one RHR Service Water subsystem becomes inoperable, subsystem is made or found to the operable subsystem and the rea o one at o 1
e' diesel-generator required for l
l to seven days unless operation of such components operability of one pump is shall be demonstrated to be i
restored within this period, operable immediately. The 1
During such seven days all operable subsystem (excluding active components of the other diesel generators) shall be RHR Service Water subsystem, demonstrated to be operable l
and its associated diesel-cenerator required for dai)y thereafter.
6peration of such components (if no external source of power were available), shall be operable.
5.
If the requirements of 3.5.C cannot be met, an orderly shutdown of the reactor shall be initiated and the reactor shall be in the Cold Shutdown
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Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Amendment No. 108 3.5-5
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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 subsystem 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, l
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
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i out-of-service time is chosen for this equipment.
If one loop I
l is out-of-service, or one pump in each loop is out-of-service, reactor operation is permitted for seven days with daily I
l testing of the operable loop (s) after testing the appropriate l diesel generator (s).
With components or subsystems out-of-service, overall core and containment cooling reliability is maintained by demonstrating the operabililty of the remaining cooling equipment.
The degree of operabililty to be demonstrated depends on the nature of the reason for the out-of-service equipment.
For 6
routine out-of-service periods caused by preventative l
l 3.5-17 Amendment No. 108
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DAEC-1 maintenance, etc., the pump and valve operability checks will be performed to demonstrate operability of the re:aaining components.
However, if a failure, design deficiency, etc., caused the out-of-service period, then the demonstration of operability should be thorough enough to assure that a similar problem does not exist on the remaining comnonents.
For example, if an out-of-service period were caused by failure of a pump to deliver rated capacity, the other pumps of this type might be subjected to a capacity
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fee [.
l[anYbi$E, ti$vNTli$cY. hrs $dUr'e"s7 N%i[tIYdd $y k'
Section6o'fthosespecificaEions,detailiherequiredextentof testing.
The pump capacity test is a comparison of measured pump performance parameters to shop performance tests.
Tests during normal operation will be performed by measuring the flow indication and/o: the pump discharge pressure will be measured and its power requirement will be used to establish flov at that pressure.
I Analyses were performed to determine the minimum required flow 3
rate of the RHR Service Water pumps in order to meet the design basis case (Reference 4) and the NUREG-0783 requirements (Reference 5).
(See Section 3.7.A.1 Bases for a discussion of the NUREG requirements.) The results of these analyses justify reducing the required flowrate to 2040 gpm per pump, a 15%
reduction in the original 2400 gpm per pump requirement.
D.
HPCI System The HPCI system is provided to assure that the reactor core is adequately cooled to limit fuel clad temperature in the event of a small break in the nuclear system and loss-of-coolant, which 3.5-18 Amendment No. 108
.c; DAEC-1
3.5 REFERENCES
1.
Jacobs, I.M., " Guidelines for Determining Safe Test Intervals and Repair Times for Engineered Safeguards", General Electric Company, APED, April
.-.w er :Pil968'I( APED ~ 57969 " WOW *wWA.- '.V-M.W" 2 e. #.'. 5Mi W. r 2.
General Electric Company, General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFR50, Appendix X, NEDO-20566, 1974, and letter MFN-255-77 from uarrell (i. Eisenhut, NRC, to E.D. Fuller, GE, Occumentation of the Reanalysis Results for the toss-of-Coolant Accident (LOCA) of Lead and Non-lead Plants, dated June 30, 1977.
4 l
3.
General Electric, Loss-of-Coolant Accident Analysis Report for Duane ArnoldEnergyCenter(LeadPlant),NED0-21085-02-1A,Rev.2, June 1982.
4.
General Electric Company, Analysis of Reduced RHR Service Water Flow at the Duane Arnold Energy Center, NEDE-30051-P, January 1983.
I 5.
General Electric Company, Duane Arnold Energy Center Suppression Pool Temperature Response, NEDC-22082-P, March 1982.
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i 3.5-26 Amendment No. 108
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DAEC-1 j
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_ LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.7 PLANT CONTAINMENT SYSTEMS 4.7 PLANT CONTAINMENT SYSTEMS Acolicability:
Apolicability:
Applies to the operating Applies to the primary and status of the primary and secondary containment system secondary containment systems, integrity.
Objective:
Objective:
:
- S y'y the integrity"of the
. To verif
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- -)To. assure. the integrity. of the
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- pyg g.d skbn*&"* * k V
~ 'p M'~sibbndaff containment systems.
containments.
Soecification:
Specification:
A.
At any time that the nuclear 1.a. The pressure suppression pool system is pressurized above water level and temperature atmospheric or work is being shall be checked once per day, done which has the potential to drain the vessel, the
- b. Whenever there is indication of suppression pool water volume relief valve operation or and temperature shall be testing which adds heat to the maintained with the following suppression pool, the pool
-limits.
temperature shall be continually monitored and also a.
Maximum water volume - 61,500 observed and logged every 5 cubic feet minutes until the heat addition is terminated.
b.
Minimum water volume - 58,900 cubic feet
- c. Whenever there is indication of relief valve operation with the c.
Maximum water temperature temperature of the suppression pool reaching 160F or more and (1)
During normal power the primary coolant pressure operation - 95F.
greater than 200 psig, an external visual examination of (2)
During testing which the suppression chamber shall adds heat to the be conducted before resumina suppression pool, the power operation.
water temperature shall i
not exceed 10F above the
- d. A visual inspection of the normal power operation suppression chamber interior, limit specified in (1) including water line regions, above.
In connection shall be made at each major with such testing, the refueling outage.
pool temperature must be reduced to below the normal power operation limit specified in (1) above withe a 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
3.7-1 AmendmentNo.108l l
DAEC-1 LIMITING CONDITION FOR OPERATION SURVEILLANCE RE0UIREMENT (3)
The reactor shall be
-scrammed from any operating condition if the pool temperature reaches 110F.
Power operation shall not be resumed until the pool temperature is reduced below the normal power operation limit specified in (1) above.
m{4).e During. reactor isolation
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conditions,thi" reactor"' "'W -
c u-shall be depressurized to less than 200 psig at normal cooldown rates if the pool temoerature reaches 120*F.
2.
Primary containment integrity 2.
The primary containment shall be maintained at all integrity shall be demonstrated times when the rehetor is as follows:
critical or when the temperature is above 212*F and a.
Type A Test fuel is in the reactor vessel excent while performing Primary Reactor Containment low power physics tests at Integrated Leakage Rate Test atmospheric pressure at power levels not to exceed 5 Mw(t).
1)
The interior surfaces of the drywell and torus shall be visually inspected each operating cycle for evidence of deterioration.
In addition, the external surfaces of the torus below.the water level shall be inspected on a routine basis for evidence of torus corrosion or leakage.
Except for the initial Type A test, all Type A tests shall be performed without any preliminary leak detection surveys and leak repairs inrnediately prior to the test.
l If a Type A test is completed but the acceptance criteria of Specification 4.7.A.2.a.(9) is not satisfied and repairs are necessary, the Type A test need not be repeated provided locally measured leakage reductions, achieved by repairs, reduce the containment's overall measured leakage rate sufficiently to meet the acceptance criteria.
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Amendment No. 108
OAEC-1 2.
There is no sign ficant thermal stratification in the condensation oscillation regime after LOCA with three feet submergence.
3.
There is some thermal stratification in the chugging regime for all break sizes.
However, this will not inhibit the pressure suppression function of the suppression pool.
4.
Seismic induced wave: 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.
Maximurn post-LOCA drawdown will not cause downcomer vent uncovering and condensation effectiveness of the suppression pool will be maintained.
Therefore, with respect to downcomer submergence, this specification is adequate. The maximum temperature at the end of blowdown tested during the Htsnbolt Bay and Bodega Bay tests was 170*F and this is conservatively taken to be the limit for complete ':ondensation of the reactor coolant, although
- condensation would occur for temperatures above 170*F.
Using a 50*F rise (Table 6.2-1, UFSAR) in the suppression chamber water 3
temperature and a minimum water volume of 58,900 ft, the 170* temperature which is used for complete condensation would be approached only if the i
3.7 32 Amendment No. 108
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DAEC-1
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1 suppression pool teeperature is 120*F prior to the DBA-l.0CA. Maintaining a pool tenperature of 95'F will assure that the 170*F limit is not approached.
As part of the progran to reduce the loads on BWR containments, the NRC l
issued NUREG-0783, which limits local suppression pool temperatures during Safety Relief Valve (SRV) actuations.
Stable stese condensation is assured in the vicinity of T-type quenchers on SRV discharge lines if the following limits on local suppression pool temperatures are met:
For all plant transients involving SRV operations during which the 1.
2 steam flux through the quencher perforations exceeds 94 lbm/ft _
the suppression pool local temperature shall not exceed soci 200*F.
1, I
2.
For all plant transients involving SRV operations during which the steam flux through the quencher perforations is less than 42 lbm/ft -sec, the suppression pool local temperature shall be at 2
least 20*F subcooled.
For all plant transients involving SRV operations during which the 3.
2 steam flux through the quencher perforations exceeds 42 lbm/ft,
sec, but less than 94 lbm/ft -sec, the suppression pool local 2
tenperature is obtained by linearly interpolating the local temperatures established under aforementioned items 1 and 2.
1 3.7-32a Amendment No. 108
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DAEC-1 t
Maintaining the suppression pool temperature below the nomal operating limit of 95'F, and scraming the reactor if the pool temperature reaches 110*F,' wf11 ensure that the local taperature limits outlined above are not exceeded during plant transients.U)
Should it be necessary to drain the suppression chamber, this should only be done idien there is no requiranant for core standby cooling systems operability as explained in Basis 3.5.G or the requirments of Spectfication 3.5.G.4 are met.
2.
Inerting Safety Guide No. 7 assimptions for metal-water reactions result in hydrogen concentrations in excess of the Safety l
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3-7-32b AmendmentNo.10q
n DAEC-1 l
1 3.7.A & 4.7.A REFERENCES l.
Section 14.6 of tne FSAR.
2.
ASME Boiler and Pressure Vessel Code, Nuclear Vessels,Section III, maximum allowable internal pressure is 62 psig.
3.
Staff Safety Evaluation of DAEC, USAEC, Directorate of
' Licensing, January 23, 1973.
4.
10 CFR 50.54, Appendix J, Reactor Containment Testing Requirements, Federal Register, August 27, 1971.
5.
DAEC Short-Term Program Plant Unique Analysis, NUTECH Doc.
No. 10W-01-065, August 1976.
6.
Supplement to DAEC Short-Term Program Plant Unique Analysis, NUTECH Doc. No. 10W-01-071, October 1976.
7.
General Electric Company, Duane Arnold Energy Center Suporession Pool Temoerature Response, NEOC-22082-P, Maren 1982.
t 3.7-49 Amendment No.108
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