ML20073T312
| ML20073T312 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 04/29/1983 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20073T310 | List: |
| References | |
| NUDOCS 8305110242 | |
| Download: ML20073T312 (5) | |
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UNITED STATES y
g NUCLEAR REGULATORY COMMISSION j
wAsHmorcN, D. C. 20666
'...../SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 89 TO LICENSE NO. DPR-49 j
IOWA ELECTRIC LIGHT AND POWER COMPANY CENTRAL IOWA POWER COOPERATIVE CORN BELT POWER COOPERATIVE l
DOCKET NO. 50-331 1
DUANE ARNOLD ENERGY CENTER s
1.0 Introduction
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The low-low set (LLS) relief logic modification for BWRs with Mark l Containments is designed to prevent multiple subsequent actuations of safety relief valves (SRV) which might normally be expected during a transient.
This in turn will reduce or prevent the discharge loads on the containment and suppression pool structures resulting from sub-sequent SRV actuations.
The discharge loads.from subsequent actuations tend to be higher due to the condensation of trapped steam in the safety relief valve discharge line (SRVDL), which results in a higher water leg in the SRVDL, and hence, larger thrust loads on subsequent actua tions.
The LLS design modification is an automatic SRV actuation system which, upon initiation, will assign preset opening and closing setpoints to two preselected SRVs.
These setpoints are selected such that the LLS con-trolled SRVs will stay open longer, thus releasing more steam (energy) to the suppression pool, and hen' e more energy (and time) will be re -
c quired for repressurization and subsequent SRV openings.
The LLS in-creases the time between (or prevents) subsequent actuations sufficiently to allow the high water leg created from the initial SRV opening to re-turn to or below its normal water level, thus, reducing thrust loads from subsequent actuations to within their design limits.
In addition, since L
the LLS is designed to limit SRV subsequent actuations to one valve, torus loads will also be reduced.
The lower MSIV water level trip causes the MSIV closure actuation to be changed from a reactor water level two signal' to a reactor water level one signal.
This de' sign modification maintains the main condenser availability for a longer time which allows more energy to be released to the main condenser and will result in a slower repressurization rate.
The lower MSIV water level trip reduces isolations, SRV challenges and provides some benefit to SRV subsequent actuations.
In a letter dated,fiarch 10, 1983 from Iowa Electric' Light and Power Company (licensee) Technical Specification Changes were requested that incorporate these logic changes into the DAEC Mark I Containment design.
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2.0 Evaluation 2.1 ' System Transient and Accident Performance and-Ov'erall plant Safety Aspects The General Ele'ctric (GE) generic evaluation submittals (Refs. L & 2) considered abnormal operational transients, design. basis accidents and the anticipated transient without scram (ATHS) events to determine the impact of these desion modifications on overall plant safety margins.
The safety evaluation for abnormal operational transients included the following considerations to determine that the design modifications will not produce any adverse' effects on safe plant systems operation and plant safety margins:
(1)
Redaction in Minimum Critical Power Ratio (MCPRl; (21 Increase in Peak Pressure; (3) Increase in Radiation Release; (4) Cause for Equipment Damage; (5) Reduction in Plant Shutdown Capability; and (6) Decrease in Core Cooling Capability.
The limiting transient events, such as MSIV trip with flux scram, and turbine trip from high power without bypass were evaluated.
General Electiic concluded that the LLS will not affect the MCPR or peak pressure, because these conditions occur early in the transients before the reactor pressure response is affected by the LLS SRVs. subsequent actuations.
The effects of the LLS on LOCAs were evaluated using the approved GE Appendix K evaluation models for the entire break spectrum.
The evaluation showed that the LLS logic has no effect on the limiting Maximum Average Planar Linear Heat Generation Rate (MAPLHGR), because the rapid reactor depressurization precludes the actuation of LLS SRVs during the design basis accident.
For ATWS events, General Electric determined that the LLS logic has no effects because this logic would not affect the reactor pressure until after the ATHS-associated short-term pressure transient is over.
It was also concluded that lowering the MSIV water level trip will not have any effects on the limiting MCPR, peak pressure and MAPLHGR during abnormal operational transients, LOCAs and ATWS events.
The impact of these design modifications on other effects such as, increase in radiation release, cause for equipment damage, reduction in plant shut-down capabili.ty, pool heat-up and decrease in core cooling capability were also considered in this generic evaluation.
The General Electric plant specific submittal (Ref. 3) identified that only non ADS-SRVs are used in the LLS logic system to reduce subsequent plant transients and that the LLS logic would extend SRV subsequent actuation time sufficiently to clear the water columns in the SRV discharge line.
Based on our review of both the General Electric generic and plant specific submittals, we have con:luded that these design modifications are acceptable, l
since they will not adversely af.fect overall plant performance and safety I
considerations.
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l-4 2.2 Low-Low Set (LLS) Circuitry 4
The LLS initiation circuitry consists of two independent and redundant channels, each of which controls power to a different SRV solenoid.
There are six SRVs at Duane Arnold, four of which are actuated by the Automatic Depressurization System (ADS).
The-two non-ADS-SRVs (B21-F0318 i
and F031F) will be used for the LLS function.
Each of the two LLS controlled SRVs will open when their respective solenoid becomes energized.
In order for either LLS channel to. energize its solenoid, both an arming logic and -
j an initiation logic.must be satisfied.
The arming logic is satisfied d
when any SRV has opened and reactor pressure (two-out-of-two logic for each i
LLS channel)'has exceeded the high pressure setpoint (this setpoint is selected above the reactor. protection system high reactor pressure scram setting to assure that a scram has occurred). Four separate reactor high pressure channels (two for each LLS channel) are used.
These instrument channels are part of the existing nuclear boiler instrumentation that provide inputs to the reactor protection system (RPS).
Inputs to the LLS arming circuitry from these channels are through isolation relays so that indepen-l dence of the RPS is maintained.
Once the arming logic for.either LLS 4
channel is satisfied, it is sealed in and annunciated in the control rocm, and remains sealed in until manually reset by the operator.
In addition, the arming logic in either LLS channel will seal in the arming logic in the other LLS channel provided the reactor high pressure permissive in that channel is satisfied.
Separation between LLS channels for this arming signal is provided by coil-contact separation.
Initial SRV actuations are detected by pressure switches located in the SRV discharge lines.
These pressure switches are set slightly above the' normal pressure expected in the discharge line'(35 psig).
Once armed, the LLS act0ation/ control logic uses nuclear boiler system re-actor pressure instrumentation to control the LLS SRV solenoids, thus opening and closing these SRVs at their assigned LLS setpoints.
This control logic 1
remains in effect as long as the arming logic is sealed in.
l Both LLS logic channels can be tested at power.
Test status lights in the control room indicate when the arming logic and control logic relays have operated satisfactorily during testing.
These test lights can also be used to I
verify proper operation of LLS seal in and reset circuits..The SRV discharge line pressure switches (three switches per SRV) and reactor pressure in-strument channels.(used both for the arming logic premissive and LLS SRV. control) must be tested separately.
Each LLS channel provides annunciation in the con-trol room' upm lost of power.
Test switches are provided to verify operability of this power monttor function.
Additional status lights are provided in the control room to indicate that the two-out-of-two reactor high pressure permissives have been satisfied. The licensee has indicated that the LLS logic will be tested consistent with the test interval for the ADS:
this was not included in the proposed Technical Specification changes.
However, the licensee in its
, letter dated April 21, 1983 committed to incorporate into the Technical Specifications within four months of the date of issuance of this amendment. surveillance testing re-quirements on the'LLS logic circuitry consistent with the test-interval for the ADS.
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4-j In addition, the licensee has initiated surveillance procedures consistent i
with the Technical Specification changes that will be submitted. We find that the proposed test frequency and commitment to incorporate this testing require-
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ment into the Technical Specifications within four months of the date of issuance of this amendment acceptable.
Power to each of the two LLS channels may be provided by one of two separate sources.
Nprmally, LLS channel "A" will be powered from 125 Vdc Distribution l
Panel No.1.
If this supply should fail, power will automatically be supplieG from 125 Vdc Distribution Panel No. 24 Similarly, LLS Channel "B" normalls -
powered from Panel No. 2 will be automatically powered from Panel No.1,'if '
d power from Panel No. 2 is interrupted.
Each LLS channel contains circuitry l
(consisting of one relay and four associated contacts to disconnect the normal supply if it fails and connect the backup supply in its place) to perform this automatic transfer function.
Since this automatic transfer feature is part of the existing protection system (ADS) circuitry and redundant protective devices (fuses and circuit breakers) are located between the automatic transfer and the safety buses, we find their design to be acceptable.
The LLS circuitry contains no channel or operating bypasses.
The circuitry added for the LLS function is located in the ADS cabinet (in a back panel room i
next to the control room) and is' separated in accordance with IEEE 384-1974.
The components of the LLS system (including power supplies) are classified as class 1E.
The LLS will remain operable in the event of loss of offsite power.
LLS components located inside the drywell are qualified for the environmental l
condition associated with a small break LOCA.
Based on our review of the General Electric generic submittals, we have con-cluded that the LLS circuitry is designed in accordance with the requirements of IEEE Standard 279-1971 to perform its intended function given a single failure and that no single failure could be found which could cause a spurious SRV actuation.
Therefore, we find the LLS logic as designed, acceptable.
2.3 Evaluation Summary System Transient and Accident Performance Overall Plant Safety Aspects.
We find these design modifications, LLS logic and lowered MSIV. water level trip, acceptable because they will not adversely affect plant performance or safety margins.
These modifications are compatible with normal plant operation and other safety systems.
l LLS Circuitry Based on our review, we have determined that the LLS modification installed at Duane Arnold is designed to perform its intended function given a single failure.
In addition, no single failure in the electrical circuits could be found which would cause a spurious SRV actuation.
The LLS is designed in accordance with the requirements of IEEE Standard 279-1971 and therefore, is acceptable.
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We have reviewed the system transients and accident performance and overall plant safety aspects and LLS circuitry related issues submitted, and we find, j
based on the above, that the design modifications including the necessary I
changes to the Technical Specifications are acceptable.
d 3.0 Environmental Consideration i
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We have determined that the amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in 'any significant envirorsnental impact. Having made this determination, we have further concluded that the amendment l
involves an action which is insignificant from the stand l
environmental impact and, pursuant to 10 CFR 551.5(d)(4) point of
, that an j
environmental impact statement, or negative declaration and environ-
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mental impact appraisal need not be prepared in connection with the j
issuance of this amendment.
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4.0 Conclusion We have concluded, based on the considerations discussed above, that:
i (1) because the amendment does not involve a significant increase in i
the probability or consequences of an accident previously evaluated, does not create the possibility of an accident of a type different from any evaluated previously, and does not involve a significant j
reduction in a margin of safety, the amendment does not involve a i
significant hazards corisideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by i
operation in the proposed manner, and (3) such activities will be l
conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
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5.0 References i
1.
General Electric letter MFN-176-82 dated November 19, 1982,
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2.
General ElectriCReport NEDE-22223 dated September 1982
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3.
General Electric Report NEDE-30021 dated January 1983.
Da ted: April 29,1983 i
Principal Contributor:
K. Desai R. Kendall
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