ML20023C228
| ML20023C228 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 05/04/1983 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20023C227 | List: |
| References | |
| NUDOCS 8305160341 | |
| Download: ML20023C228 (5) | |
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UNITED STATES
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g NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555
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. COOPER NUCLEAR STATION l.0 Introduction Tha low-low set (LLS) relief logic modification for BWRs with Mark 1 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 (SRVOL), which.results in a higher water leg in the SRVDL, and hence, larger thrust loads on subsequent actuations.
~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 set trolled SRVs will stay open' points are selected such that tne LLS con-longer, thus releasing more steam (energy) to the suppression pool, and heni:e more energy (and time) will be re-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 subsaquent actuations to within their design 1imits.
In addition, since the LLS is designed to limit SRV subsequent actuations to one valve, torus loads will also be reduced.
l The lower MSIV water level trip causes the MSIV' closure actuation to be l
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::A820003, Application for Amend to License DPR-46 Changing Tech Specs, Safety Relief Valve (S/Rv) Low-Low Set (LLS) Sys & Lower MSIV Water Level Trip. Encl Withheld (Ref 10CFR2.790)|letter dated December 17, 1982]] from Nebraska Public Power District (licensee)
Technical Specification changes were requested that incorporate these logic changes into the CNS Mark I Containment design.
This was supplemented by information provided in a February 15, 1983 letter.
l 8305160341 830504 PDR ADOCK 05000298 P
. 2.0 Evaluation 2.1
' System Transient and A'ccident Performance and-Overall Plant Safety Ascects Tha General Electric (GE) generic evaluation submittals (Refs. L & 2) considered abnormal operational transients, design basis accidents and the anticipated transient without scram (ATWS) events to detemine 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 )
Reduction in Minimum Crittcal Power Ratio (MCPR).;
(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 Electric concluded that the LLS will not affect the MCPR or peak pressure, because these condit:fons occur early in the transients before the reactor pressure response is affected by the LLS SRVs subsequent actuations.
Tha effects of the' LLS on LOCAs were evaluated using the approved GE Appendix X evaluation models for the entire b'reak spectrum.
The evaluation showed that the LL$ 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 ATMS events, General Electric dotermined that the LLS logic has no effects l
because this logic would not affect the reactor pressure until after the ATWS-associated short-term pressure tr,ansient 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 l
also considered in this generic evaluation.
l The General Electric plant specific submittal (Ref. 3) identified that only l
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 I
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 cortluded that these design modifications are acceptable, since they will not adversely affect overall plant performance and safety I
considerations.
. l I
2.2 Low-Low Set (LLS) Circuitry The LLS initiation circuitry consists of two redundant channels each of which controls power to a different SRV solenoid.
There are eight SRVs at CNS, six of which are actuated by the Automatic Depressurization System (ADS).
The two non-ADS SRVs (71D and 71F) 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 an initiation logic must be satisfied. The arming logic is satisfied when any SRV has opened and reactor pressure (two-out-of-two logic for each 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 provides inputs to the reactor protection system (RPS).
Once the arming logic for either LLS channel is satisfied, it is sealed in and annunciated in the control room, and remains sealed in until manually reset by i
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 ch'annel is satisfied.
Separation between LLS channels for this arming signal is provided by coil-contact separation.
Initial SRV actuations aro detected by pressure switches located in the SRV discharge lines.
These pressure switches are set slightly above the normal pressure expected in the dis-charge line (30 psig). Ohce armed, the LLS actuation / control logic uses nuclear boiler system reactor pres'sure instrumentation to control the LLS SRV solenoids, L
thus opening and closing these SRVs at their assigned LLS setpoints. This control logic remains in effect as long as the arming logic is sealed in.
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 verify proper operation of LLS seal in and reset circuits.
The SRV discharge line pressure switches (one switch par SRV) and reactor pressure instrument channels (used both for the arming logic psrmissive and LLS SRV control) must be tested separately.
Each LLS channel provides annunciation in the control room upon loss of power.
Test switches are provided to verify operability of this power monitor function.
Additional status lights are provided in the control room to indicate that the two-out-of-two re-actor high pressure permissives have been satisfied.
The licensee has included in their proposed plant Technical Specifications changes monthly testing of the pressure instrumentation used for LLS SRV control, and semiannual testing of th2 LLS logic.
Technical Specification surveillance requirements already exist for the SRV discharge line pressure switches and the reactor pressure channels used for the LLS arming logic permissives.
These test frequencies are acceptable to the staff.
Power to each of the two LLS channels may be provided by one of two separate sources.
Normally, both LLS channels will be powered from the Division 1 125 Vdc supply.
If this supply should fail, power will automatically be supplied to both LLS channels from the Division 2125 Vdc supply.
Each LLS channel contains circuitry (consisting of one relay and four associated contacts to disconnect the Division 1 supply if it fails and connect the Division 2 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 since redundant protect.ive devices (fuses and circuit breakers) are located between the auto-matic transfer and the safety buses, the staff finds this design to be acceptable.
. The LLS circuitry contains no channel or operating bypasses.
The circuitry added for this LLS function 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 conditions associated with a small break LOCA.
Based on our review, we have determined that the proposed LLS modification to be in-stalled at CNS is designed to perform its inter.ded 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 l
in accordance with the requirements of IEEE standard 279-1971 and there-fore, is acceptable.
j 2.3 Evaluation Summary System Transient and Accident performance Overall plant Safety Aspects.
W2 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.
LLS Circuitry.
Based on our review, we have determined that the LLS. modification installed at Cooper Nuclear Station is designed to perform its intended function given e single failure.
In addition, no single failure in the electrical circuits could b3 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.
3.0 Environmental Consideration W] have determined that the amendment does not authorize a change in affluent types or total amounts nor an increase in power level and will not result in any significant environmental impact.
Having made this detemination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR 151.5(d)(4), that an l
environmental impact statenent, or negative declaration and environ-mental impact appraisal need not be prepared in connection with the issuance of this amendment.
4.0 Conclusion l
We have concluded, based on the considerations discussed above, that:
(1) because the amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated, does not create the possibility of an accident of a type different
5-from any evaluated previously, and does not involve a significant reduction in a margin of safety, the amendment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety af the public will not be endangered by operation in the proposed manner, and (3) such activities will be 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.
5.0 References 1.
General Electri,c letter MFN-176-82 dated November 19, 1982, 2.
General ElectriCReport NEDE-22223 dated September 1982.
3.
General Electric Report NEDE-22197-dated December 1982.
Dated:
tiay 4,1983 Principal Contributor:
K. Desai R. Kendall e
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