ML20077D383
| ML20077D383 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 07/20/1983 |
| From: | IES UTILITIES INC., (FORMERLY IOWA ELECTRIC LIGHT |
| To: | |
| Shared Package | |
| ML112230892 | List: |
| References | |
| NUDOCS 8307260501 | |
| Download: ML20077D383 (10) | |
Text
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PROPOSED CHANGE RTS-152 TO THE-DUANE ARN0LD ENERGY CENTER TECHNICAL SPECIFICATIONS The holders of License DPR-49 for the Duane Arnold Energy Center propose to amend Appendix A (Technical Specifications) to said license by deleting current _pages and replacing them with the attached new pages. A list of the affected pages is given below.
The purpose of this change is to reduce the required minimum flow rate of the Residual Heal Removal Service Water (RHRSW) pumps as given in Section 4.5.C.1(b) of the Technical Specifications. This change was prompted by numerous instances of failing to meet the required flow rate during surveillance testing.
As part of the effort to solve the problem, Iowa Electric contracted General Electric (GE) to analyze the RHRSW system to determine the minimum flow rate required to meet the design basis conditions. The RHRSW system's primary function is to provide cooling water to the Residual Heat Removal (RHR) system heat exchangers during various modes of the RHR system. The design specification for the RHR system states that the shutdown cooling mode is considered to be the d W gn basis, but that the steam condensing mode should also oe evaluated at it may sometimes govern. Attachment 1 contains the GE analysh of these mtdes of the RHR system. The results of which justify a reduction of up to 30% in the RHRSW flow rate from the present requirements.
As part of the Mark I containment modifications, suppression pool temperature limits were set which would prevent unstable steam condensation during blowdowns to the containment. NUREG-0783 established guidelines for demonstrating conformance to the limits on local suppression pool temperatures for T-type quenchers on Safety Relief Valve discharge lines. contains the GE analysis on the suppression pool cooling mode of the RHR system using the NUREG guidelines.
Appendix B of this report demonstrates conformance to the above limits with a 15% reduction in the present RHRSW flow rate.
The design bases for modes of the RHR system which use the RHR heat exchangers have been analyzed with reduced RHRSW flow. The results of these analyses show that a 15% reduction in the presently required RHRSW flow rate is justified.
The changes being made are as follows:
1)
Reduce by 15% the required minimum flow rate for the RHRSW system as given in Section 4.5.C.1(b).
2)
Update the Bases and References for Section 4.5 to support the 15%
reduction in RHRSW flow.
%M O
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4,
l 3)
Update the Bases and References for Section 3.7 to include a discussion of the NUREG-0783 requirements and the results of the GE analysis.
4)
Consolidate text.on pages 3.7-1, 3.7-la, 3.7-2 and delete pages 3.7-la and 3.7-lb.
List of Pages Affected 3.5-5 3.5-18 3.5-26 3.7-1 3.7-la*
3.7-lb*
3.7-2 3.7-32 3.7-32a**
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3.7-49
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DAEC-1 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT Item Frequency b)
Flow Rate After major Test-Each pump RHR service maintenance water pump and every 3 shall deliver months at least 2040 l
gpm at a TDH of 610 ft. or more.
2.
From and after the date that 2.
When 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 reason, reactor operation must and the other subsystems shall be limited to thirty days be demonstrated to be operable unless operability of that immediately 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.
From and after the date that 3.
When one RHR Service Water I
one RHR Service Water subsystem becomes inoperable, Subsystem is made or found to the operable subsystem and the be inoperable for any reason, diesel-generators required for reactor operation is limited operation of such components to seven days unless shall be demonstrated to be operability of that subsystem operable immediately and daily is restored within this thereafter.
period. During such seven days all active components of the other RHR Service Water subsystem and its associated.
diesel-generators required for operation of such components (if no external source of power were available), shall be operable.
3.5-5
DAEC-1 maintenance, etc., the pump and valve operability checks will be performed to demonstrate operability of the remaining components.
However, if a failure, design deficiency, etc., caused +be out-of-service period, then the demonstration of operability should be thorough enough to assure that a similar problem does not exist v.
the remaining components. 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 test.
In any event, surveillance procedures, as required by Section 6 of these specifications, detail the required extent of 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/or the pump discharge pressure will be measured and its power requirement will be used to establish flow at that pressure.
Analyses were performed to determine the minimum required flow 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%
l reduction in the original 2400 gpm per pump requirement.
D.
HPCI System I
The HPCI system is provided to assure that the reactor core is adequately cooled to limit fuel clad temperature in the event o#
1 a
small break in the nuclear system and loss-of-coolant, which 3.5-18
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 1968 (APED 5736).
2.
General Electric Company, General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFR50, Appendix K, NED0-20566, 1974, and letter NFN-2bS-77 from Uarrell G. Eisenhut, NRC, to E.D. Fuller, GE, Documentation of the Reanalysis Results for the Loss-of-Coolant Accident (LOCA) of Lead and Non-lead Plants, dated June 30, 1977.
3.
General Electric, Loss-of-Coolant Accident Analysis Report for Duane Arnold Energy Center (Lead Plant), NED0-21082-02-1A, Rev. 2, June 1982.
4.
General Electric Company, Analysis of Reduced RHR Service Water Flow at the Ouane Arnold Energy Center, NEDE-30051-P, January 1983.
I 5.
General Electric Cr.apany, Duane Arnold Energy Center Suppression Pool Temperature Response, NEDC-22082-P, March 1982.
I 3.5-26
DAEC-1 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.7 PLANT CONTAINMENT SYSTEMS 4.7 PLANT CONTAINMENT SYSTEMS Applicability:
Applicability:
Applies to the operating Applies to the primary and status of the primary and secondary containment system secondary containment systems.
integrity.
Objective:
Objective:
To assure the integrity of the To verify the integrity of the primary and secondary primary and secondary containment systems.
containments.
Specification:
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 witer volume - 58,900 cubic feet
- c. Whenever there is indication of relief valve operation with the c.
Maximum water temperature temperature of the suppression I
pool reaching lf0F or more and (1)
During normal power the primary coolant pressure operation - 95F.
greater than 200 psig, an external visual examination of l
f (2)
During testing which the suppression chamber shall adds heat to the be conducted before resumina suppression pool, the power operation.
water temperature shall 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 i
normal power operation limit specified in (1) above within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
3.7-1 l
a
DAEC-1 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT (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.
(4)'
During reactor isolation conditions, the reactor shall be depressurized to less than 200 psig at normal cooldown rates if the pool temperature reaches 120*F.
2.
Primary containment integrity 2.
The primary containment shall be maintained at all integrity shall be demonstrated times when the reactor is as follows:
critical or when the temperature is above 212*F and a.
Type A Test fuel is in the reactor vessel except 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 innediately prior to the test.
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 contain:nent's overall measured leakage rate sufficiently to meet the acceptance criteria.
3.7-2 l
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q r-DAEC-1 to downcomer submergence, this specification is adequate. The maximum 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.
I Using a 50*F rise (Table 6.2-1, UFSAR) in the suppression chamoer 3
watt.r temperature and a minimum water volume of 58,900 ft, the 170*F temperature which is used for complete condensation would be approached only if the suppression pool temperature is 120*F prior to the DBA-LOCA. Maintaining a pool temperature of 95'F will assure that the 170*F limit is_not approached.
As part of the program to reduce the loads on BWR containments, the NRC. issued NUREG-0783, which limits local suppression pool temperatures during Safety Relief Valve (SRV) actuations. Stable steam 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:
1.
For all plant transients involving SRV operations during which the steam flux througn tne quencher perforations exceeds 94 lbm/ft -sec, the suppression pool local 2
temperature shall not exceed 200*F.
I 3.7-32
l DAEC-1 2.
For all plant transients involving SRV operations during which the steam flux through the quencher perforations 2
is less than 42 lbm/ft -sec, the suppression pool local.
temperature shall be at least 20*F subcooled.
3.
For all plant transients involving SRV operations during which the steam flux through the quencher perforations 2
exceeds 42 lbm/ftz-sec, but.less than 94 lbm/ft -sec, the suppression pool local temperature is obtained by linearly interpolating the local temperatures established under aforementioned items 1 and 2.
Maintaining the suppression pool temperature below the normal operating limit of 95'F, and scramming the reactor if the pool temperature reaches 110*F, will ensure that the local temperature limits outlined above are not exceeded during plant transients.(7)
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 Bas., 3.5.G or the requirements of Specification 3.5.G.4 are met.
I 2.
Inerting Safety. Guide No. 7 assumptions for metal-water reactions result in hydrogen concentrations in excess of the Safety 3.'7-32 a k_.
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DAEC-1 3.7.A & 4.7.A REFERENCES 1.
Section 14.6 of the 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 UAEC Short-Term Program Plant Unique Analysis, NUTECH Uoc. No. 10W-01-071, October 1976.
7.
General Electric Company, Duane Arnold Energy Center Suppression Pool Temperature Response, NEUC-22082-P, March 1982.
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