ML20079L746
| ML20079L746 | |
| Person / Time | |
|---|---|
| Site: | 05000605 |
| Issue date: | 11/01/1991 |
| From: | Marriott P GENERAL ELECTRIC CO. |
| To: | Pierson R Office of Nuclear Reactor Regulation |
| References | |
| MFN-138-91, NUDOCS 9111070083 | |
| Download: ML20079L746 (14) | |
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Gt ikc! var Energy Nosember 1,1991 NIFN No.138 91 I)ocket No. STN 50 605 til!N 9178 Document Conti01 Desk U.S. Nuclear itegulatory Commission Washington.11C. 20$$5 Attention: Itobert C. Pierson, Director Standardization and Non Power Reactor l'ioject Directorate
Subject:
Advanced inniling Water ( AllWit) Technical Specillrations (TS) lleference: Advanced lloiling Water (AllWR) Technical Specifications (TS),
Chester Posiusny (NRC' >o Patrick W. hiatriott (Gli), htFN No.
103 91, dated Septembei 12,1991.
As requesttd in the reference letter, Gli has reviewed the AllWit design relatoe to llWit/6 (and earlier designs, where appsopriatn) to determine the pertinent differences with regards to the technical specifications ('IS) currently envisioned for the AllWit design. The enclosed document summariies the major design dif ferences (Table I), presents an assessment of how the TS diller from the improved TS (Table 2), provides a summary of what is new, dif ferent or not applicable (Tab;c 3) and an example of how the '15 would be written where they dif fer from the improved 'IS (Attachment 1). Similur TS writeups willis issued in the near future (date to be determined in the upcoming November 8 meeting) for the res of the areas that are dillerent. We hope that this document will help facilitate the NitC review and acceptance at the complete set of Allwit TS, which will closely follow the ongoing improved 'IS ettort.
Sincerely, l
P.W. hi$rriott, hianager Regulatory and Analysis Services hi/C 382, (408) 925 6948 cc: F. A. Ross (Doll)
N. D. Fletcher (DOE)
C. Poslusny, Jr. (NitC)
R. C. Iler lund J.F.Ouir
( 0 11)
(GE) }0/
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tm AD00: 039Cy[ ]/)h La,,-.
i AllWR Tecimical Specifications (Th l
The AllWR design has been reviewed relative to llWR/6 (and cailler designs, where appropriate) to determine the pertinent dilTerences with regaids to the technical specifications currently envisioned for the design.
Table 1 sununarites the major design difreiences identined as being within the scope of technical specifications reguliements, a'ul thus possibly necessitating a change to the draft improved technical speellications (ITS) currently being finalized by the NRC aint the llWR Owner's Gioup. l Table 2 presents a comprehensive assessment of how the technit specifications envialoned for AllWR will differ from the draf t ..a. With the exception of one section, the assessment was made iciative to the llWR/6 ITS.
For Section 3.6, Contaimnent Systems, the llWR/4 ITS was utilised, as the layout of this section more closely paralleled that anticipated for AllWR. In the table each of the individual system sperincations from the ITS have been listed t and categorlied in 1 of 3 ways iciative to its application to AllWR, The three -
categories are: not applicable (NA): essentially the same (ES); and. design results in different specification (DI)). Ex all NA and DD listings, as well as some ESldanatory listings. conunents
'he ES category includes are included items where there is absolutely no change as well as items where although the design may differ slightly, the technical sperlucation requirements (and wording) are fundamentally the same. '
Table 3 provides a more concise sununary of what is new, different or not applicable for AllWR. Also included is the one new specification envisioned for AllWR dealing with the essential multiplexing system (EMS). The EMS is used in conunon by many of the critical instrumentation systems and therefore is considered desening of a separate specification to assure its operability.
Attachment I provides some indication of how the AllWR speci0 cations will be written, where the icquirements differ from those in the draft ITS. Inchides an example of proposed specifications for control rod drive accunmlators and 1he reactor iccirculation system. These specifications are the only DD items from Sections 3.1 and 3.4, respectively, of the ITS. Please note that the llases (br these two specifications are not included, instead, a brief explanation is provided after each item to indicate why the AllWR TS differs from the ITSs in each case, in general, the full AllWR liases for each sperincation will closely follow the corresponding ITS liases, modified to account for the differenences explanation punided.
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- TABLE 2 - .
. LIMITING CONDITIONS FOR OPERATION (LCOs) AND SURVEILLANCE PEQUIREN_%'TS4SRs)
Sv: tion Description Category Coerments 3.1 REACTTbiTY CONTROL SYSTEMS 3.1.1 SHUTDOWN MARGIN (SDM) ES Rod
reactivity egiaivalent of highest worth rod pair 3.1.2 Reactivity Anomalies .ES 3.1.3 Control Rod OPERABILITY ' E5 P.od Pairs j .s.1.4 Control Rod Scram Times ES Rod position measured in percentage insertion rather than in notches 3.1.5 Control Rod Scram Aw, Nators DD No Rx Pressure Scram Assist, Rods are scrammed by accumulator pressure only; Auto scram on low charging water header pressure; rod pairs 4 3.1.6 Rod l'attern Control 'ES 3.1.7 Standby Liquid Control (SLC1 System ES No explosive valves, equivalent otherwise 3.1.P Scram Disel;arge Volume (SDV) V+2nt and Drai1 Valves NA TJo Scram Discha .;e Volume 3.2 POWER DISTRIBUTION UMITS 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE ES i (APLH3R) 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR) EE
- 3.2.3 LINEAR HEAT GENERATION RATE (LHGR) lapp!icable ES to Non-GE Fuel Only) j 3.2.4 Avecage Pc,wer Range. Monitor (APRM) Gain and ES S ooints !
'I DD - Dessrt D#ferer.ces ' ES - Essentially the Same f4A - tJot Appiche 11/1'91 E-
.l TABLE 2 - LCOs and SRs (contirated) ,
3.3 WSTRUMENTATION 3.3.1.1 Reactor Protection System (RPS) Instrumentation DD Digitally multiplexed system; 2 of 4 sensor logic input,2 of 5 divisional trip logic output; spec to be more similar to CE ITS than past BWRs 3.3.1.2. Source Range Monitor (SRlW) Instrumentation DD : Wide range SRNMs replace SRMs and IRMs; SRNMs provide fast period RPS trip 3.3.2.1 Control Rod Riock lastrumentation DD Some basic inputs; monitoring and initiation of
- . blocks implemented it. slightly different ways 3.3.3.1 Post-Accident Monitoring (PAM) Instrumentation ES 3.3.3.2 Remote Shutdown System ES 3.3.4.1 End-of-Cycle Recirculation Pump Trip'(EOC-RPT) OD Digitally multiplexed system,2 of 4 logic
. Instrumentation 3.3.4.2 Anticipated Transient Without Scram-Recirculation DD Digitally multiplexed system,2 of 4 logic l
Pump Trip (ATWS-RPT) Instrumentation 'i 3.3.5.1 Emergency Core Cooling System (ECCS) DD Actuation Logic is 2 of 4 on input,2 of 2 on j instrumentation output 1 3.3.5.2 Reactor Core Isolation Cooling (RCIC) Instrumentation NA Part of ECCS spec above 3.3.6.1 Primary Containment isolation (PCI) Instrumentation- DD Digitally multiplexed system,2 of 4 logic 3.3.6.2 Secondary Containment ! solation (SCI) Instrumentation DD Digitally multiplexed system,2 of 4 logic 3.3.6.3 Containment Spray System Instrumentation NA No Auto actuation on ABWR 3.3.6.4 Suppression Pool Makeup System (SPMS) NA System not part of ABWR instrumentation 3.3.6.5 Relief and Low-Low Set (LLS) Instrumentation NA No LLS function on ABWR; Credit not needed for relief mode of actuation 3.3.7.1 Control Room Fresh Air'(CRFA) Instrumentation DD Digitally multiplexed system,2 of 4 logic 3.3.8.1 Loss of Power (LOP) instrumentation ES 3.3.8.2 RPS Electrical Power Monitoring (EPM) NA Not part of ABWR design 3.3.8.3 Load Shedding and Sequencing (LSS) instrumentation ES DD - Design Dtterences ES - EssereJany the Same f4A - Not Appleable 11/1/91 4-i 5
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TABLE 2 - LCOs and SRs (continued)
Description Category Comments Section 3.4 REACTOR COOLANT SYSTEM -(RCS)
Recirculation Loops Operating DD 10 Rx Internal pumps vs. 2 external loops 3.4.1 NA Not part of ABWR design 3.4.2 Flow Control Valves (FCV)
Jet Pumps NA Not part of ABWR design 3.4.3 ES Ref. BWR/4 - Safety Mode Only 3.4.4 Safety / Relief Valves (S/RVs) -
RCS Operational LEAKAGE ES Minor differences due to LBB 3.4.5 RCS Pressure isolation Valve (PIV) Leakage ES 3.4.6 RCS LEAKAGE Detection Instrumentation ES 3.4.7 RCS Specific Activity ES 3.4.8 1 ES 3 subsystems available, only 2 required 3.4.9 Residual Heat Removal (RHR)-Shutdown RCS Pressure and Temperature (P/T) Limits ES No loop-to-loop differential tcmperature pump 3.4.10 start requirement I ES 3.4.11 Reactor Steam Dome Pressure 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOUNG (RCIC)
SYSTEM DD 6 suosystems: 3 completely sep_. rate divisions, j 3.5.1 ECCS-Operating each with both a high pr. and a low pr.
subsystem; mcreased redundancy provides greater leeway for initial subsystem out of service ES Same basic requirements 3.5.2 ECCS-Shutdown NA Combined with ECCS - Operating 3.5.3 RCIC System PJA f401 Apo:cde DD - Desgn D.tferences ES - Esser 8ah the Same 4
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.6 iTAllLE 2 - LCOs and SRs (continued) ,
l Section Description Category Comments 3.6 CONTAINMENT SYSTEMS -This section based on BWIV4 IT3- ,
3.6.1.1 Primary Containment ES 3.6.1.2 Primary Containment Air Locks ES 2 airlocks for ABWR, same otherwise
, 3.6.1.3" Primary Containment isolation Valves (PCIVs) ES ;
3.6.1.4 Primary Containment Pressure ES i 3.6.1.5 Primary Containment Air Temperature ES 3.6.1.6 Suppression Chamber-to-Drywell Vacuum Breakers ES ,
3.6.1.7 Reactor Building-to-Suppression Chamber Vacuum NA Not part of ABWR design Breakers 3.6.1.8 Low Low Set (LLS) Safety / Relief Valves (S/RVs) 'NA Not part of ABWR design i 3.6.1.9 MSIV Leakage Control System (LCS) NA Not part of ABWR design Suppression Pool Average Temperature SRNM equivalent for IRM indication of 1% CTP f l 3.6.2.1 ES f
3.6.2.2 Suppression Pool Water Level ES 3.6.2.3 Residual Heat Removal (RHR) Suppression Pool Cooling DD 3 subsystems available and required; allows System extended interval for initial subsystem O.O.S.
3.S.2.4 Residual Heat Removal (RHR) Suppression Pool Spray ES f 3.6.2.5 Drywell-to-Suppression Chamber Differential Pressure NA Mair.tenance of differential pressure not required in ABWR i 3.6.3.1 Primary Containment Hydrogen Recombiner System ES is permanently installed in ABWR l
(PCHRS)- MODES 1 and 2 (if permanently installed)
! 3.6.3.2 Primary Containment Hydrogen Mixing System (HMS) NA ,
' lot part of ABWR design f l - MODES 1 and 2 ,
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3.6.3.3 Primary Containment Oxygen Concentration ES l
l 3.6.3.4 Containment Atmosphere Dilution (CAD) System NA Not part of ABWR design j
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I DD - Design Ddferences ES - Essentialty the Same NA - Not Appreable r i
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TABLE 2 - LCOs and SRs (continued) .
Section Description Category Comments 3.6.4.1 Secondary Containment ES 3.6.4.2 Secondary Containment Isolation Valves (SCIVs) ES 3.6.4.3 Standby Gas Treatment System (SGTS) ES Common charcoal filter train; if inop, makes both subsystems inop 3.7 PLANTSYSTEMS 3.7.1 IStandbyl Service Water (SSW) System and Ultimate DD Each division of ECCS (incl. EDGs and RHR decay Heat Sink (UHS) heat removal) has own dedicated cooling water and service water subsystemsr although design differs, approach will be same 3.7.2 High Pressure Core Spray (HPCS) Service Water System NA Part of overall safety related cooling and service (SWS) water systems spec above 3.7.3 Control Room Air intake, Recircu!ation, and Purification ES (AIRP) System 3.7.4 Control Room Heating, Ventilation and Air Conditsor ng ES (HVAC) System 3.7. 5 Main Condenser Offgas ES 3.7.6 Main Turbine Bypass System ES 3.7.7 Fuel Pool Water Level ES t
DD - Design Differences ES - Essentalty the Same N A - Not Appicable 11/1/9 t 4
TABLE 2 - LCOs and-SRs-(continued) D 4
Section Description Category Comments -
3.8 ELECTRICAL POWER SYSTEMS !
3.8.1 AC Sources.- Operating DD 3 EDGs and 2 Offsite sources available (one . l
- offsite source is via backfeed thru gen. output !
- breaker); ABWR design configuration results - '
in different LCO and AOTs, especially for
.' initial equipment out of service #
t 3.8.2 AC Sources - Shutdown ES Although design differs, spec is similar (written . ,
-am supported system perspective) 3.8.3 Diesel Fuel and Lubricating Oil ES DC Sources '- Operating 3.8.4 DD 4 divisions of DC in ABWR t l' 3.8.5 DC Sources - Shutdown E5 Although design differs, spec is similar (written from supported system perspective) ;
3.8.6 Battery Electrolyte' ES I
, 3.8 7 inverters - Operating DD 4 DivisionalInverters Fed from 4 DC and 3 AC .
Divisions: Basic Approach is Similar, but some !
. special cases likely ,
3.8.8 inverters : . Shutdown ES Although design differs, spec is similar (written -;
, from supported system perspective) 3.8 9 Distribution System - Operating DD 3 AC divisions,4 DC divisions; No division i dedicated to HPCS: Division 4 DC may be special case l r
t 3.8.10 Distribution System Shutdown ES Although design differs, spec is similar (written from supported system perspective) j
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DD - Design Odferences ES - Essedial!y the Same NA- Fbt Appicable
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. TABLE 2 - LCOs and SRs (continued) h CJetegory h 3.9 REFUEUNG SYSTEMS
- . 3.9.1 Refueling Equipment Interlocks' ES 3.9.2 Refuel Position One-Rod-Out interlock ES Scram testing done by HCU rod pair 3.9.3 Control Rod Position ES 3.9.4 Control Rod Position Indication ES 3.9.5 Control Rod OPERABILITY'- Refueling ES Test is via step rather than notch I 3.9.6 Reactor Pressure Vessel (RPV) Water Level ES 3.9.7 Residual Heat Removal lRHR)- High Water Level ES 1 of 3 subsystems ren'd 3.9.8 Residual Heat Removal (RHR)- Low Water Level ES 2 of 3 subsystems req'd 3.10 SPECIAL OPERATIONS 3.10.1 Inservice Leak and Hydrostatic (ISLH) Testing Operation ES 3.10.2- Reactor Mode. Switch interlock Testing ES 3.10.3 Single Control Rod Withdrawal - Hot Shutdown ES CRDs scram tested in pairs; SDM analysis-assumes / supports this
- 3.10.4 Single Control Rod Withdrawal - Cold Shutdown ES CRDs scram tested in pairs; SDM analysis assumes / supports this 3.10.5 Single Control Rod Drive (CRD) Removal - Refueling ES CRDs scram tested in pairs; SDM analysis assumes / supports this 3.10.6 Multiple Control Rod Withdrawal - Refueling ES 3.10.7 Control Rod Testing - Operating ES 3.10.8 Shutdown Margin (SDM) Test - MODE 5 ES 3.10.9 Recirculation Loops - Testing ES Although design differs, similar relief is needed and appropriate 3.10.10 Training Startups ES DD - Design Deferences ES - Essenda3y the Same NA - fJot Apphcaole 11/1/91 4- i i
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. TAllt.E 3 -
Summary of ITS Itetns New, Different or Not Applicable for AllWR Specific!ttigos Not Acolicable (NA) to ABWR 3.1.8 Scram Discharge Volume (SDV) Vent and Drain Valves 3.3.5.2 Reactor Core kolation Cooling (RCIC) Instrumentation 3.3.6.3 Containment Spray System Instrumentation 3.3.6.4 Suppression Pool Makeup System (SPMS) instrumentation 3.3.6.5 Relief and Low Low Set (LLS) instrumentation 3.3.8.2 Reactor Protection System (RPS) Electrical Power Monitoring (EPM) 3.4.2 Flow Control Valves (FCV) 3.4.3 - Jet Pumps 3.5.3 RCIC System 3.6.1.7 Reactor Building to Suppression Chamber Vacuum Breakers 3.6.1.8 Low Low Set (LLS) Safety / Relief Valves (S/RVs) 3.6.1.9 Main Steam isolation Valve (MSIV) Leakage Control System (LCS) 3.6.2.5 Drywell to Suppressiors Chamber Differential Presecte 3.6.3.2 Primary Containment Hydrogen Mixing System (HMS)- MODES 1 and 2 3.6.3,4 Containment Atmosphere Dilution (CAD) System 3.7.2 High Pressure Core Spray (HPCS) Servico Water System (SWS)
Soecifications Where Qgslan is Different_(DD) for ABWR 3.1.5 Control Rod Scram Accumulators 3.3.1.1 Reactor Protection System (RPS) Instrumentation 3.3.1.2 Source Range Monitor (SRM) Instrumentation ,
3.3.2.1 Control Rod Block Instrumentation 3.3.4.1 End of Cycle Recirculation Pump Trip (EOC RPT) Instrumentation 3.3.4.2 Anticipated Transient Without Scram Rncirculation Pump Trip (ABVS RPT)
Instrumentation 3.3,5.1 Emergency Core Cooling System (ECCS) instrumentation 3.3.6.1 Primary Containment isolation (PCI) instrumentation 3.3.6.2 Secondary Containn.ent Isolation (SCI) Instrumentation 3.3,7.1 Control Room Fresh Air (CRFA) Instrumentation 3.3.8.1 Loss of Power (LOP) Instrumentation 3.4.1 Recirculation Loops Operating 3.5.1 ECCS-Operating 3.6.2.3 Residual Heat Removal (RHR) Suppression Pool Cooling System 3.7.1 IStandbyl Service Water (SSW) System and Ultimate Heat Sink (UHS) 3.8.1 AC Sources - Operating -
3.8~4 DC Ecurces - Operating-3.8 7 Inverters _ Operating
-3.8.9 Distribution System - Operating Etoposed New Specincations for ABWR 3.3.x x Essential Multiplexing System 11/1/91 4
ATTACHMENT 1 3.1 REACTMTY CONTROL SYSTEM 3.1.5 CDntrol Rod Semm Accumulators LCO 3.1.5 All control rod scram accumulators shall be OPERABLE.
APPLICAB;LITY: MODES 1 and 2.
ACTIONS l
.........................y O T E .........................
l Separate Condition entry is allowed for each control rod scram accumulator j CCNDITION REQUIRED ACTION COMPLETION TIME A. One or more control rod scram A.1 Declare the associated control 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> accumulators inoperable rod or control rod pair inoperable.
SURVEll1ANCE REQUIREMEt#S SURVEILLANCE FREGENCY SR 3.1.5.1 Verify control rod scram accumulator pressure is > [ 1850 ] 7 days psig.
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ATTACHMENT 1 (continued)
Explanallon for Olfferences in ABWR Specification Relative to Drall ITS The ABWR CRD system differs from past BWR designs in several key respects. With regards to the toch spec for control rod scram accumulator operability the portinent differences aru that each HCU provides the scram force for a pair of control rods (escept for the contor control rod which has its own accumulator) and because there is no scram dischargo volumo the scram function must be accomplished agamst ioactor pressure (i.e. thoro is no scram assist from reactor pressure). Thus, a properly charged accumulator is the only means for assuring scram. To protect against the simultaneous loss of adequato pressure in niu!tiple scram accumulators (such as would occur on loss of CRD charging water duo to the trip of the running CRD pump) an automatic scram is initiated by the RPS system on censed few CRO charging water heador pressure. This assures a scram occurs while titero is still sufficient accumulator chargo to scram all rods.
The ABWR spec for this function la based on the traditional BWR spec for the case with low I RPV pressure. However, there is no nood to make the spec conditional on lo^w charging water heador pressure as there is an automatic ca.fety function in place to address this concern.
Consequently, affected rods are simply declared inoperable within a short t;mo. This otsures thct they are fully inserted and disarmed, via the spec or1 control rod operabilitr. cuch thc the accumulator function is no longer needed. ,
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, , ATTACHMENT 1 (continued) 3.4.1 Reactor Internal Pumos (RIPS) Operating L C O 3.4.1 At least nine RIPS shall be in operation.
APPLICABILITY: MODES 1 and 2, ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One required RIP not in A.1 ........N O T E..........
operation. Provisions of LCO 3.0.4 are not applicable.
Reduce THERMAL POWER to s I hour 95% RTP.
B. Two required RlPs not in B.1 ........N O T E..........
operation. Provisions of LCO 3.0.4 are not applicable.
Reduce THERMAL POWER to $ 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 90% RTP, C Three or four required RIPS C.1 noduce THERMAL POWER to s 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> c.ot in operation. 25% RTP.
AND C.2 Restore at least seven RIPS to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> from operation. Initial discovery of less than seven RIPS in operation.
D Five or more required RIPS D.1 RedalTHERMAL POWER to s 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> not in operation. 5% RTP.
AND D.2 Restore at least seven RIPS to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> from.
operation, initial discovery of less than seven RIPS in operation E Required Action and E.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time not met.
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- ATTACHMENT 1 (continued)
SURVEILLANCE REQU REMENTS SURVEILLANCE FREQUEtCY
,SR 3.4.1.1 Verify at least nine RIPS are in operation. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Explanation for Oliferences in ABWR Specification Relative to Draft ITS The ABWR has 10 recirculation pumps internal to tho vossol rather than two recirc pumps located on external loops as with recent 'BWRs. However, the bases for this LCO is essentially the same as in the past, i.e. the operating state of the recirculation system must be consistent with the power / flow operating conditions assumed in the plant LOCA and transient analyses.
For ABWR rated core flow can be attained with only nine of ten RIPS in operation. However, the core flow that can be attained with less than nine RIPS operating is less. Thoroforo, at least nine RIPS are required to be in operation to ensure during a LOCA the assumptions of the LOCA analysis are satisfied without restriction. With less than nino RIPS in operation, all potential power and flow operating states have not been accounted for in either the LOCA or transient analysis. Therefore, certain restrictions apply depending on the number of RIPS operating, With less than nine RIPS in operation the THERMAL POWER must be restricted so that the assumptions of the LOCA and transient analyses are met. With only seven or eight RIPS operating THERMAL POWER is restricted to 5 90% and s 95% RTP, respectively, However, operation may continue indefinitely. Also, as noted, LCO 3.0.4 is not applicable for these conditions. With less than seven pumps operating,1HERMAL POWER is restricted even further and operation may only continue for a short timo.
For the case of 5 or 6 pumps running, THERMAL POWER must be reduced to s 25% RTP because of potential stability concerns. With less than 5 pumps operating, power must be reduced to s 5% RTP due to the lack of detailed analysis of the actual flow distribution with less than half of the pumps in operation providing forced flow at higher power levels.
With less than seven RIPS operating the steady stato power and flow characteristics of the core have not been fully analyzed. Therefore, even at reduced power levels, continued operation is allowed for only a short time while an attempt is mado to restoro at least seven pumps to operating status. With less than seven pumps restored to operating status within the Required Completion Time, the reactor is required to be in MODE 3. In this condition, the RIPS are not required to be operating because of the reduced severity of design basis accidents and minimal dependence on the forced flow characteristics.
Because of the increased number of pumps involved, and their distribution around the periphery of the bottom head region, pump to pump flow mismatch is not of concern for the ABWR. Thus, the surveillance requirernent specified simply checks for pumps in opera %n, and not for flow mismatch.
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