ML20205P529
| ML20205P529 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 11/04/1988 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20205P526 | List: |
| References | |
| NUDOCS 8811080197 | |
| Download: ML20205P529 (7) | |
Text
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S "ETY EVALUATION BY T,HE_0FFICE OF NUCLEAR RE,A,CT,0R_R,EGULATI,0,N.
E GPU NUCLEAR CORPORATION AND JFRSEY CENTR,AL, P,0,WE,R_ A _ LIGHT _C,0PfANY OYSTER CR_E_EK NUCLEAR GENERATING STATION COMPLI_ANAE WITH ATWS RULE 10 CFR 50.62 RELATING TO ALTERNATE RCD INJECT _ ION (ARI) AND RECIRCULA_ TION PUFP,(,RP,T), S,Y,5,T, EMS DOCKET NO. 50-219 1.0,1N,T,RODUCTION On July 26, 1984, the Code of Federal Regulations (CFR) was amended to include Section 10 CFR 50.62, "Requiren.ents for Reduction of Risk froc Anticipated Transients Without Scram (ATWS) Events for Light-Water-Cooled Nuclear Power Plants" (known as the "ATWS Rule"). An ATWS is an expected operational transient (such as loss of feedwater, loss of condenser vacuum, or loss of offsite pcwer) which is accon:panied by a failure of the reactor trip system (RTS) to shutdown the reactor. The ATVS rule requires specific in.provements in the design and operation of comercial nuclear power facilities to reduce the likelihood of failure to shutdown the reactor following anticipated transients, and to niitigate the consequences of an ATWS event.
For each boiling kater reactor, three systems are required to mitigate the consequences of an ATKS event.
I.
It must have an alternate rod injection (ARI) system that is diverse (from the RTS) from sensor output to the final achation device. The ARI system must have redundant scram air header exhaust valves. The ARI system must be designed to perform its function in a reliable manner and be independent (from the existing RTS) from sensor output to the final actuation device.
2.
It must have a standby liquid control system (SLCS) with a minimum flow capacity and boron content equivalent in control capacity to 86 gallons per minute of 13 percent by weight of sodium pentaborate solution. The SLCS and its injection location must be designed to perform its function in a reliable manner.
3.
It must heve equipment to trip the reactor coolant recirculating purrps autorraticelly under conditions indicative of an ATKS.
This equiptrent must be desigwd to perfmn its function in a reliable manner.
h!b00$ 0 0 9
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This Safety (Evaluation addresses the ARI system (Item 1) and the ATWS/
RPT system Item 3). The SLCS (Item 2) was addressed in our Safety Evaluations dated February 18, 1988 and August 3,1988.
2.0 REVIEW CRI,T,E,R_IA The systems and equipment required by 10 CFR 50.62 do not have to meet all of the stringent requirernents nonna11y applied to safety-related equipment. However, this equipment is part of the broader class of structures, systems, and corrponents impor tant tc safety defined in the introduction to 10 CFR 50 Part 50, Appendix A, Gennal Design Criteria (GDC). GDC-1 requires that "structures, systems and corraonents important to safety shall be designed, fabricated, erected and tes'ed to quality standards comensurate with the importance of the safety functions to be performed." Generic Letter 85-06 "Quality Assurance Guidance for ATWS Equipment that is not Safety Related" details the quality assurance that must be applied to this equipment.
In general, the equipment to be installed in accordance with the ATWS Rule is required to be diverse fron. the existing RTS, and must be testable at power.
This equipitent is intended to provide needed diversity (where only minirnal diversity currently exists in the RTS) to reduce the potential for connon mode failures that could result in an ATWS leading to unacceptable plant conditions.
The criteria used in evaluating the licensee's submittal include 10 CFR 50.62, "Rule Considerations Regarding Systems and Equipment Criteria" published in Federal Register Volurre 49, No.124 dated June 26, 1984 and Generic Letter 85-06. "Duality Assurance Guidance for ATWS Equipirent that is r et Safety Related."
3.0 OYCTER CREEK Nu', LEAR GENERATING STATION ARI & RPT JJ,ST,E,M_D ETCYIPTIM'~ ~" - ~ ' " ~ " " ~
GPU Nuclear Corporation (GPUN/ licensee) has provided information by letters dated Septerrber 3,1987 December 30, 1987 and April 29, 1988 that describes the equipment utilized to mitigate the potential consequences of an ATWS without scrani event.
The systens consist of reactor pressure and reactor water level sensors, logic, power supplies, and instrumentation to initiate the protective actions to mitigate an ATWS event. The protectise actions are to include:
a.
Alternate Rod Injection (ARI) b Recirculation Purnp' Trip (RPT)
The 1).nsee referenced the BWR Owners' Group (BWROG) topical n eport NEDE-31096-P.
"Res
.e to NRC ATWS Rule 10 CFR 50.62" as the basis for confonr.ance to the ATWS Rule
' API and RPT systems are independent from the RTS and are capable of initi-
active actions during any loss of offsite power event. The lic*n...... to install the ARI system during the 1988 refueling outage.
The ARI cynem uscs a two-out-of-two logic which will cause the irrrrediate energization of the ARI valves when either the reactor vessel high pressure trip setpoint or the reactor vessel low-low water level trip setpoint is reached, or the manual pushbutton, located in the control roca, is actuated.
The ARI valves and bleed paths are sized to allow injection of all rods to l
begin within 33 seconds and all rod trotion completed within 43 seconds, i
l
The function of the RPT is to re6uce the severity of thernal transients on fuel elenents by tripping the recirculation pumps early in the transient events (such as turbine trip, or load rejections). The rapid core flow reduction increases void content and thereby introduces negative reactivity in the reactor to reduce the thernal power. Oyster Creek installed the RPT system in 1980.
The RPT is a "Modified Hatch Logic" using a one-out-of-two taken twice circuit.
There are two separate and independent circuits to trip the recirculation pumps.
The redundant sensors detect high reactor pressure or low reactor water level.
The Oyster Creek Nuclear Generating Station design has two independent trip coils in each drive motor breaker for each reactor recirculation pump.
The RPT end ARI systems can be tewed while the plant is operating (excluding purrp breakersandARIvalves). This test will check the systems from the sensor outputs thru the logic to the actuating devices.
The ARI function can be reset by tranual activation of the ARI reset switches after a 43 second tire delay.
This delay ensures that the ARI scraq will go to completion.
4.0 EVALVATI0N OF ARI SYSTEM 4.1,1RJ,,Sy,s, tem Fun, tion Time By letter dated Cecen.ber 30, 1907, the licensee stated that bastd on the scram discharge volone fill time calculations, activation of all vent paths will illow the insertion of all control rods to begin within 33 seconds and be conpleted within 43 seconds.
The staff finds that these times are not the same as approved by the SER to the topical report NECE-31096-P wh'ch requires that rod inhetion motion will begin within 15 seconds and be completed within 75 seconds f ron, ARI initiation.
The licensee is required to provide a plant unique study to denonstrate that all the ARI objectives identified in the BWROG Topical Report can still be met with the longer ARI initiation tin.e span.
The licensee is also required to perform a preoperational test to verify that actual ARI function time will not be longer than the calculated tines. The licensee has verbally agreed to perforra both the analysis and test. We will confirm completion of both items during the staff inspection of the ATKS Rule implementation (in accordance with temporary instruction 2500/20 dated Febnfary 9,1987).
4.2 3 a f e ty-R e 1 a 1,e d,R e q,u,1,r,ep,e n,t,slI E ELS,t,a n,d a,r,d,,2J,9)
The ATWS Rule does not require the ARI system to be safety grade, but thc implen.entation nust be such that the existing protection system t.ontinbes to treet all applicable safety-related criteria.
The licensee identified the ARI systera as a ncn-1E system.
The ARI system interfaces with the safety-related systems through qualified isolation devices. The API system does not degrade existing design requirements for the RTS and ECCS This is in conformance with the ATKS Rule guidarce and therefore is acceptable.
4.3 Redundancy The ATVS Rule requirss that the ARI system must have redundant scram air header exhaust valves, but the ARI system itself does not need to be returdant.
0.
The Oyster Creek Nuclear Generating Station ARI system has redundant header exhaust valves. The ARI performs a function redundant to the backup scram system. This is in confonrance with the ATVS Rule guidance and therefore is acceptable.
4.4 DiversityFromExistinglTS.
The licensee stated that the instrument component 5 of the ARI system are diverse from existing RTS canponents with the exception of the reactor low-low level trip channels. The reactor low-low level trip function utilizes Rcsemount Transmitters and Foxboro SPEC 200 r.odules for instruirent loop signal processing. The RTS also uses the sane type of signal processing equipment.
The ARI system corrponents are required to be diverse from the RTS from sensor output to the final actuation device as stated in the ATWS Rule. Diversity was the most fitportant factor regarding the implen,entation of the ATWS mitigation equipment because corroon mode failures were determined to be a larger safety risk than random failures. Based on the relative irnportance of this diversity requirement, the staff has concluded that the type of signal processing equipment provided for the Oyster Creek ARI system is not acceptable in that it is identical to the.TS signal conditioning equipment.
o Having considered the need for additicnal time to change out the existirg trip units to provide con.pliance with 10 CFR 50.62, the Comission agrees with an extension of tine to fully comply with 10CFR50.62 until no later than the end of the next refueling cutage (Refueling Outage 13R). Another alterr.ative, permitted by the provisions o+ 10 CFR 50.12, wculd be to request an exemption fron. the diversity requirement of 10 CFR 50.62.
However, we do not reconcerd this option.
4.5 Electric _al Indep,endence From The Existing,,R,T,5 The ATWS Rule guidance states that the ARI systen is required to be electrically independent from the existing RTS ' rom sensor output to the final actuation device at which point ncn-safety related circuits must be isolated frcm safety-related circuits.
The ARI system is identified as a non-safety related system. The pressure input signals come from non-safety related Appendix R fuel zone pressure instrurrents.
The level input signals cane from safety-related ECCS reactor vessel low-low water level instruments. The level signals are isolated from the class 1E signal within the Foxboro cabinets through qualified isolation devices. The ARI system circuits are isolated from safety-related circuits through qualified isolation devices (relay to contact isolation). The ARI logic and valve control power is separated from the RTS control power. This i: in confonr.ance tdth the AT./S Rule guidance and therefore is acceptable.
4.6
,P,hy,s,1,ca,1,,S, epa,ra,t,1,o,n, From Existing RTS The ATWS Rule guidance states that the implementation of the ARI system trust be such that separation criteria applied to the existing protecticn systen, are not violated.
I
. The ARI system from sensor outputs, transmitters, trip units and associated circuits are separated and independent from the RTS. A physical barrier will be provided between the ARI and the safety related circuits where required to meet the plant separation criteria. The separation between the RTS and the ARI system is identified as satisfying the guidance provided in the design of the safety systems for this plant. The staff finds this acceptable.
4.7 Enyj,r,o nme n,t a l,,qu,aj i fi c a t i o n The ATWS Rule guidance states that the qualification of the ARI system is for anticipatet operational occurrences only, not for accidents.
The ARI system equipment will be located outside the drywell which is a non-harsh environc4ent during an ATWS event. The ARI system is qualified to operational occurrence condition. The staff finds this acceptable.
4.E,C,ualj,tylssurance NRC Generic Letter 85-06 dated April 16, 1985 3rovides q.ality assurance guidance for the API system. The licensee has stated tlat this guidance will be followed.
The staff finds this acceptable.
4.9,Sa,f,e,ty,,Re,l a,t e,d,,( I E ) Pow e r S u p pl y The ATWS Rule guidance states that the ARI system must be capable of performing its safety function with loss of offsite power, and that the power source should be independent from the existing RTS. The ARI system control logic and valves are powered from a non-class IE DC bus which is independent from the RTS pcher sources.
The DC buses are backed-up by batteries not affected by a loss of offsite power. The licensee has stated that a fonnal FMEA is available to denonstrate that no action of either the ARI or RTS can affect the performance of the other system. The /PI system is capable of performing its safety functions with loss of offsite power and the ARI control pcwer scurces are independent from the existing RTS power source. This design is in coaforrance with the ATWS Rule guidance and therefore is acceptable.
4.10,T,es,ta,bj lj tyy _At Power t
The ATkS Rule guidance states that the ARI system should be testable at power.
The Oyster Creek ARI system (excluding the API valves) can be tested at power.
A Normal / Bypass Switch is used to bypass ARI logic for testing purposes during plant operation. A status and availability condition will be annunciated. The ARI system uses a two-out-of-two for level or pressure logic arrangen,ent.
Each individual level and pressure instrument can be tested during plant operation by disabling one trip relay to avoid initiating the ARI system since either two low level or two high pressure signals will initiate the action. This is in conformance with the ATWS Rule guidance and therefore is acceptable.
4.11,I n a,dy,e,r,t,e,n,t,,A,c t u a t i o n The ATWS Rule guidance states that inadvertent ARI actuation which challenges other safety systems should be minimized.
a.
d
- E.
The ARI system actuation setpoints, reactor 1cw-low water level or reactor high pressure, were chosen such that nomal reactor scram would occur prior to the ARI function. The ARI system is energized to actuate and is initiated when either two Reactor Low Water Level signals or two Reactor High Pressure signals are present.
This is in conformance with the ATWS Rule guidance cod therefore is acceptable.
4.12 Manual Initiation The ARI system can be manually initiated from the control rocen.
The staff finds this acceptable.
4.13 Infomation_Re,a,dout The ARI system provides status and system availability indications in the control room. ARI status and availability annunciators which indicate system initiation and system off-ncmal are available in the control roce. This is in confornance with the ATWS Pule guidance and the,efore is acceptable.
4.14 Cco p l e t t o,n, p,f, P r o te c t i,v e A,c,t i,o n_,0 nc e,,IL,1,s, Lnit i a t ed The ART has a 43 second seal-in festure to ensure the completion of protective action once it is initiated. After initial conditions return to norral, deliberate operator action is required to reset the system logic to noma 1 This is in conforriance with the ATWS Rule guidance and therefore is acceptable.
4.15 C,o,nc,1,u,s,1,o,n _0_n ARI System As stated in Reference 3, the staff SFD on BWP0G Toaical Report NEDE-3109e-P, the staff does not intend to repeat its review of t1e design % formation described in the BWROG Topical Report. We find that it is acceptable tt reference this report in a specific license application.
Reference 1 sumarizes the licensee's compliance with the ATWS Rule.
The staff finds that one area (e.g. diversity) of the Oyster Creek ARI design is not in compliance with the ATWS Rule, 10 CFR 50.67, and as a result, the ARI system is not acceptable.
Ttis is discusstd in more detail in Section 4.4 of this evaluation.
The licensee '
also required (as a confimatory item) to perfom a preoperational test t a vs..fy that tie actual ARI function time is within the design limit.
5.0 Evaluation Of A,TES/R,P,T, Sy,s,t,em By letter dated September 3,1987, the licensee stated that the present ATWS/RPT system at Oyster Creek station consists of two trip systems and uses a roc lfied Hatch logic, a one-out-of-two twice circuit.
Either high reactor pressure or low-low reactor water level from two separate channels will energize a trip coil in each recirculating pump motor-generator set drive motor breaker.
It uses two 1cdependent coils in each breaker.
As stated in Reference 3, the modified Hatch design is an acceptable reference for the ATWS/RPT design.
The staff concludes that the Oyster Creek ATWS/RPT design is in conpliance with the ATWS Rule,10 CFR 50.62 paragraph (c)(5) and therefore is acceptable.
. +
6.0,TE,C,H_NJ,C,Aj._ SPECI FI CATIONS The equipment required by the ATWS Rule to reduce the risk associated with an ATWS event must be designed to perform its function in a reliable manner. A method acceptable to the staff for demonstrating that the equipment satisfies the reliability requirements of the ATWS Rule is to provide equipment technical specifications including Operability ar.d Sury2illance requirements. The staff will provide guidance on Technic 61 Specification requirements for the ARI system and the ATWS/RPT system til a separate docun'ent.
7.0 REFERENCES
1.
GPU Nuclear Corporation letter, R.F. Wilson to NRC Document Control Desk dated September 3, 1987.
P.
BWROG Topical Report NEDE-31096-P, "Ar.~icipated Trarisients Without Scram; c
Response to NRC ATKS Rule 10 CFR 50.67 " dated December 1985 i
3.
Staff SER on BWROG Topical Report NEDE-31096-P, letter from Gus Lainas (NRC) to Terry A. Pickens (BWR Owners' Group chairman), dated October Pl. 19b6.
4.
GPU Nuclear Corporation letter, R.F. Wilson to NRC Document Control Cesk dated December 30, 1987.
5.
OPU Nuclear Corporation letter, R.F. Wilson to NRC Document Control Desk dated April 29, 1989.
Dated:
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