ML20132F410
| ML20132F410 | |
| Person / Time | |
|---|---|
| Issue date: | 11/24/1982 |
| From: | Rubenstein L Office of Nuclear Reactor Regulation |
| To: | Novak T Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20132A685 | List:
|
| References | |
| FOIA-83-619 NUDOCS 8507180435 | |
| Download: ML20132F410 (22) | |
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e7 l m 24 maz q/yyjr(Lg , ,q;u l MEMORANDUM FOR: Thomas M. Novak, Assistant DirBctor for Licensing, Division F g
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l of Licensing, NRR , FROM: L. S. Rubenstein, Assistant Director for Core and Plan't Systems, Division of Systems Integration, NRR
SUBJECT:
DESIGN COMPARISON BETWEEN SNUPPS AND SIZEWELL "B" NUCLEAR PLANTS In response to the October 25, 1982 letter from H. Denton to Division Directors, the f.uxiliary Systems Branch and the Power Systems Branch have reviewed Item 7
" Increased Equipment Redundancy" of Enclosure 1 to the referenced letter.
The results of our review are enclosed and follow the format of Enclosure 2 to the Denton letter. In sumary we have identified the differences between the SNUPPS and Sizewell designs for the auxiliary feedwater system, component cooling water system, essential service water system and electrical distribution system. The Power Systems Branch performed the review of the electrical dis-tribution system. Also, for these systems, we described the SNUPPS design, showed how the SNUPPS design meets our criteria, and identified areas for further exploration to help in understanding the design differences. An analysis of the safety benefit due to the design differences is being per-formed by the Division of Safety Technology (DST) and will be transmitted under a separate cover letter after we have DST's input. bes & L. S. fiu enstein, Assistant Director for Core and Plant Systems Division of Systems Integratien cc: R. Mattson J. Lazevnick
- 0. Parr J. Knight R. Lobel M. Srinivasan D. Eisenhut W. LeFave
Contact:
W. LeFave X29470 I l G/f/
)l 8507180435 850606 PDR FDIA SHOLLYB3-619 PDR
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- 7. Increased Equipment Redundanc;y
- a. TITLE .
Auxiliary Feedwater System - 6esign Differences Associated With The Sizewell B Use of Four ASW Pumps Instead Of the-SNUPPS Three Pump System - try DESCRIPTION - The Sizewell B AFW system consists of four AFW pumps (2 Motor Driven, 2 Turbine Driven) that comprise two independent subsystems of the AFW system, made up of the motor driven subsystem and turbine driven subsystem. Each subsystem nomally takes suction from its, own con- . densate storage tank (each of which is the same capacity as the one CST at SNUPPS) during nomal operation and as in the case of SNUPPS, each pump is capable of discharging to any combination of 1 to 4 steam generators. However, the'Sizewell retor driven p0mps discharge to their own dedicated nozzles in the steam generators, while the turbine driven pumps discharge to the main feedwater system piping in the same fashion as all three pumps (2 motor driven, 1 turbine driven) at SNUPPS. Nomally, each pump is lined up to feed two steam generators automatically via the same initiating signals as SNUPPS. Each of the pumps at Sizewell is rated at 100 percent while at SNUPPS the motor driven pumps are each 100 percent and the turbine driven pump is 200 percent. - Each condensate storage tank at Sizewell B is protected against natural phenomena including earf.hquakes, and a backup unprotected long tem supply (dedicated to AFW) is available to each pump via manual valving. TheSNUPPSdesignhasanunprotectedCSTwith'ajrotecieTautomatic._..' _ _ . . . l long tem supply which is the essential service water system. I l S
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9 -s Another difference in the two systems is the AFW flow control valves.
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At Sizewell, the flow control valves ari semi-automatic in that
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after system initiation they will cycle full o?cn or closed' testng steam generator high and low level setpoints as inputs.. This continues until the operator intervenes to take manual control to maintain specified level. At SNUPPS the valves do not cycle open or closed but remain at a preset throttled position until the operator takes control to maintain the specified level. The final major difference in the two systems is AFW flow following a main steamline break or main feed 11ne break,that results in an unisolable steam generator. The Sizewell B includes a feed only good steam generators system similar to the US Babcock and Wilcox designs that automatically tenninate 'AFW flow to the faulted steam generator. The SNUPPS design relies on flow limiting orifices (similar to other W designs) to limit the flow to the faulted steam . generator thereby assuring sufficient AFW flow to the intact steam generators. The SNUPPS design requires operator action within 10 minutes to secure flow to the faulted steam generator. SNUPPS DESCRIPTION The SNUPPS AFW system consists of three AFW pups (2 Motor Driven, 1 Turbine Driven) taking suction from one condensate storage tank during nonnal operations, with each pump capable of discharging _,,y__,m,, _-. - - . . m ,~_.. , . . . . . _ ----~.-.- -_ __._ _
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t . l 3_ i via the main feedwater system' piping to any combination of 1 to 4 steam generators. Normally, one motorIiriven pump is " lined up to automaticallysupplywatertotwosteamgeneratorswhiletheother motor driven pump is lined up to automatically provide water to the remaining two steam generators. The turbine driven pump is normally lined up to automatically supply water to all four steam generators. U. S. ACCEPTANCE CRITERIA AND HOW THEY WERE MET BY THE SNUPPS DESIGN
- 1. GDC2 " Design Bases for Protection Against Natural Phenomena" The entire AFW system, except for the normal water supply from the CST, is designed to seismic Category I requirements, and is located in the seismic Category' I, flood , missile , and tornado protected auxiliary building. In the event of failure of the CST due to natural phenomena or other causes, the MW pump suction is automatically transferred to the essential service water (ESW) system which is protected against natural phenomena. Therefore, in the event of any natural phenomena the system would automatically
, deliver water to the steam generators and the requirements of GDC-2 are met. 1 9 e
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- 2. GDC-4 " Environmental and Missile Design Bases" - -
Each AFW pump and associatedactive valves are located in separate cubicles such that any flooding, missile, harsh envitynment or pipe break would only affect the capability of the pump train within a given cubicle which would assumed to be disabled due to the failure. i The remaining trains would be available to supply adequate flow even assuming a single failure. Additionally the AFW system is not used for startup, hot standby or shutdown and therefore is not considered a high energy system except for those portions of the steam supply line to the turbine and the portions of piping connected to the main feedwater system that are pressurized during normal plant opera-tions. Also all necessary components within the system will be environmentally qualified to operate under accident conditions, includ-ing pipe break and loss of ventilation. i
- 3. GDC-5 " Sharing of Structures, Systems and Components" 1
( The AFW system at SNUPPS is not shared between reactor units.
- 4. GDC-19 " Control Room" The AFW system is automatically initiated on receipt of an auxiliary feedwater actuation signal (AFWAS) and delivers flow to the steam gener-ator without operator action.. Steam generator water level control is then manually controlled by the operator from the control room or remote shutdown panels. The operator can monitor steam generator pressure, level and AFW flow at either station. The system therefore meets the requirements of GDC-19 regarding prompt initiation of plant shutdown.
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- 5. GDC-44 " Cooling Water" and GDC-34 " Residual Heat Removal" Adequate isolation of the AFW system from non essential systems is included in the system design. The AFW system coffhects-to-the main feedwater system in the safety-related portion of the main feedwater line downstream of the main feedwater line isolation valve.
Essential lines connecting to the condensate storage tank are provid-ed with automatic isolation valves to isolate the AFW system from the tank in the event of tank failure. These isolation features meet the isol'ation requirements of GDC-44. The AFW system can function automatically in the event of a loss i of offsite power. The heat transfer path under this condition is from the steam generators to the at'mosphere via the safety related atmos-pheric dunp valves. The turbine-driven pump receives main steam from two of the four steam generators through an air-operated, fail open valve, one on each of the two steam generators. These valves are normally closed, and open automatically on receipt of a turbine-driven pump initiation signal. Each motor-driven pump is powered from a separate diesel generator backed bus. The AFW system discharge valves are air operated and nonnally open. They also fail open on loss of normal air and are provided with a seismic Category I backup accumulator (backed up by nitrogen). Based on the above, plus the reliability analyses performed by the applicant and reviewed by the staff, the design meets GDC-34 and GDC-44 with respect to its ability to remove decay heat from the reactor under accident conditions. l J l
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e 5 j 6- , . In accordance with the single-failure criterion of GDC-34 and GDC-44 the SNUPPS AFW system is designed to accomodath a single failure in any active system component without loss of function. The AFW' system consists of three trains, supplying all four steam generators. Each train is supplied by one AFW pump. The turbine-driven pump train normally supplies all four steam generators. Each of the two motor-driven pump trains supplies two of the four steam generators. The motor-driven and turbine-driven pump trains are connected together downstream of the AFW valves before the connections to the main feed lines. . The AFW control valves for each train are powered from separate power supplies, such that a power supply failure only affects one train. The valves associated with the turbine-driven pump are all powered from a battery backed source, such that AC power is not required forturbine trair. operation. A failure modes and effects analysis for the entire AFW system results in the loss of not more than one train. Thus, adequate feedwater is ensured to at least two steam generators in the event of a pipe break or other postulated design basis accident concurrent with a single active failure. The decay heat removal requirements of GDC-44 are met since each AFW pump is designed to provide 100 percent of the flow necessary for residual heat removal over the entire range of reactor operation, including all postulated design basis accidents. e f
- 6. GDC-45 and GDC-46 " Inspection of Cooling Water Systems" and " Testing . '
of Cooling Water Systems" - r Each pump is equipped with a recirculation line to .the CST which can ; be used for periodic functional testing. Periodic testing of pumps and valves is perfomed in accordance with the Westinghouse Standard Techni-- cal Specifications. The limiting conditions for operation also follow
! the Westinghouse Standard Technical Specifications and the recomendat'ons i derived as a result of the TMI accident. The safety related portions of the system have also been designed and located in areas that are accessible during normal plant operation to pemit periodic inservice inspection.
- 7. Branchlechnical Position ASB 10-1 " Design Guidelines for Auxiliary Feedwater System Pump Drive and Power Supply Diversity for Pressur-ized Water Reactor Plants" To meet the diversity requirements the turbine-driven AFW pump train provides a diverse means of ensuring feedwater supply to the f
steam generators independent of all offsite or onsite AC power sources for at least two hours. The turbine-driven pump bearings do not require cooling from an AC dependent source, and the pump can operate without area-forced ventilation for the 2-hour period. Automatic actuation and control of this train is provided from the vital DC power source. l 8. Branch Technical Position RSB 5-1 " Design Requirements for Residual Heat Removal System" i The AFW system has the capability to perinit operation at hot shut-I down for at least four hours followed by cooldown to the RHR cut-in temperature from the control room using only safety grade equipment and assuming the worst case single active failure. This is assured by virtue of safety grade atmospheric dump valves and a safety-grade long term source of water from the ESW system. Thus the reouirements of-RSB 5-1 with respect to AFW are met. t .
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- 9. II.E.1.1 Auxiliary Feedwater System Reliability Evaluation -
The SNUPPS AFW system also meets all the recomendations of the 1NI Task Action Plan outlined in NUREGS-0660 and 073i. The manner in which these recomendations are met are sumariz5 below. -
- a. Recomendation GS The proposed SNUPPS Technical Specifications will be based on NUREG-0452. Revision'3, " Standard Technical Specifications for Westinghouse Pressurized Water Reactors". This
' meets the outage time limit and subsequent action time with >
respect to one AFW pump and its, associated flow train and i essential instrumentation being inoperable. ,
- b. Recomendation GS The SNUPPS AFW system has an automakic switchover to the ESW system such that the single manual valve from the CST cannot interrupt AFW flow and there are no other single valve closures that could in'terrupt all AFW system flow.
- c. Recomendation GS Throttling AFW flow will not be used to l avoid waterhamer at' SNUPPS. Therefore, this recomendation is not applicable.
- d. Recomendation GS Although automatic switchover to the ESN water supply is included in the design, emergency proced-ures will . . be available for transferring the AFW pump i suction to backup water supplies,.and the single manual valve in the line from the CST to the AFW system will be locked l
open by physical means such as a chain and padlock. l I 1 -
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- e. Reconnendations GS-5 and GL-3 '- Since the turbine-driven pump train is capable of being automatically initiated and controlled
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independent of any AC power source for at leas 5 two hours, these recommendations are met. 3
- f. Reconnendation GS Flow-path availability from the CST to i the steam generators via en AFW train will be confimed as part
! of functional testing on return from any extended cold shut-down. A second independent operator will also perfom valve lineup checks following return to service of a train that has been out for testing or maintenance,
- g. Reconnendation GS-7 and GL The AFW' system automatic initiation and control circuitry and signals are redundant and meet safety-I grade requirements.
I h. Reconnendation (1) - The condensate storage tank level is monit-1 ored by redundant transmitters one of which is AFW pump suction pressure and is powered from redundant Class 1E power sources. Each transmitter provides indication and alam in the control room. The low level alarm setpoint allows at least 20 minutes for operator action. These alams are also backed by the safety i ""~" ~~ ! l-relatedswItchovertoth'eESW.sys'temI' ~' T . l
- i. Reconnendation (2) - The applicant has connitted to perfoming I
i a 48-hour endurance test on each pump to assure long term operability. The results of these tests will be reviewed by the staff. I j. Reconnendation (3) - Redundant safety-grade flow instrumentation , to each steam generator with indication in the control room is pmvided. The instrument channels are powered from the emergency buses that satisfy the diversity requirements of ASB 10-1. f
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- k. Recomendation (4) - Since two. AFW flow paths are available during periodic tests, this recomendation does5not apply.
- 1. Recomendations GL-1 thru GL These recomendations either ,
do not apply or are met as described under the short term . (GS) recomendations. . Analysis - provided in Separate Section r Areas for Further Evaluation
- 1. Detemine whether decision to add another turbine-driven pump was based on single failure concern during a complete loss of all AC power, or, whether redundancy in diverse power sources was the basic concern without considerition of complete loss of AC power.
- 2. . Detemine the rationale for adding separate AFW inlet nozzles
, on the steam generator for the motor-driven pumps. Noting that it adds four containment penetrations to the design, inquire which safety concern or accident scenario was instrumental in
,the decision.
- 3. Was the feed only good steam generators system designed by
, Westinghouse, Bechtel or by others? At any" rate "has a detailed failure analysis been performed to assure that no failure modes could result in isolation of more than the one (faulted) steam generator including momentary pressure fluctuations? 8 O l
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- b. .TJT1[, ,
Design differences between the SNUPPS and Sizewell B component cooling water (CCW) systems and essential , servicewater(ESW) systems. DESCRIPTION , ,, , , , The CCW systems for both plants are basically the same with two-100 percent trains each of which has two-100 percent l pumps. The major differences are (1) the Sizewell B CCW des.ign is allocated more loads because the ESW system uses seawate'r, and (2) there-is a forc6d-air (radiator), heat exchanger that can be used to remove heat fmm the CCW system as a backup to the ESW system. The backup heat exchanger is necessary since the Sizewell B ESW system is not seismically designed. The SNUPPS CCW system has only one main heat exchanger (cooled by ESW) since it is capable of maintaining CCW temp-erature within design limits under all conditions. The Sizewell B CCW system has two heat exchangers cooled by ESW because, given the higher heat load of the Sizewell design, the main heat exchanger cannot maintain CCW water temperature low enough to cool containment air coolers, pump room coolers, and pump lube oil heat exchangers. Consequently, an auxiliary heat exchanger automatically supplements CCW cooling under accident conditions. At SNUPPS, the ESW system supplies. cooling water for many of these heat loads. An added feature at the Sizewell B plant is the addition of two reserve ultimate heat sink (RUHS) pumps that have only one functionr To provide CCW flow to the containment air coolers 1.n. the event that all CCW pumps s
are not available. The pumps.are mariually started.and; stopped and are not designed to handle full accident loads such as a t large LOCA. . The RUHS heat exchangers associated with the CCW system are automatically switched into the system on loss of ESW system flow due to conmion mode failure of the ESW pumps such as screen blockage or a seismic event. They are not designed to handle the design basis LOCA heat loads since a large LOCA concurrent with a design basis earthquake is considered beyond the design basis. Although a large LOCA concurrent with an SSE is also considered beyond the design basis for SNUPPS. the ESW system and UHS for SNUPPS is designed for the SSE. l The ESW system at Sizewell has four-100 percent pumps while the SNUPPS ESW system has two. However at Sizewell two of the pumps are necessary for normal operation.- while'SilVPPS ESW pumps are ; only used during emergencies. During nonnal operatio'ns at SNUPPS, the ESW system heat loads are supplied by two out of three (50 percent capacity each) station service water pumps. At first glance the ESW systems appear to be quite different for Sizewell and SNUPPS but further inspection shows they are very similar since each plant has two-100 percent headers with each header having two pumps capable i of supplying water with one of the two pumps normally operating. The. ESW system at Sizewell B is not designed to remain functional - following a seismic event since it is hcked up by the seismically o
- qualified RUHS. At SNUPPS the ESW system is designed to remain functional following a seismic event.
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. . .. SNVPPS CCW System isescription. -
The CCW system at SNUPPS consists of two separate 100 percent trains each capable of being supplied by two-100 percent pumps (4 pumps total). Each train has one surge tank,,and one CCW he'at exchanger
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cooled by the ESW system. _ A common nonseismic header is used to supply nonsafety-related 4 loads during normal operation. This comen loop is connected to both essential trains via automatically operated safety-grade isolation valves that close upon receipt of a safety injection . .
- signal, surge tank low level or high flow indicative of a pipe failure.
- Redundant safety related components cooled by the CCW system are the RHR heat exchangers, RHR pump seal coolers, centrifugal charging pump bearing oil coolers, SIS ' pump bearing oil coolers, and the fuel pool heat exchangers.
During normal operation one pump in one of the safety-related loops is operated with the redundant safety-related train isolated from the system. If the pump should fail the remaining pump in the same operating loop automatically starts. ' Upon loss of offsite power one pump in each of the safety-related loop's automatically startswhen sequenced onto the diesel generator ~ buses. The remaining pumps stay in auto-standby.- SN'UPPS E s Sysle[.'Descrin:.icn The ESW system at SNUPPS consists of two independent 100 percent trains, with one 100 percent pump per train which are in standby during normal plant operation. During normal operation the ESW loads are supplied cooling water by the nonsafety grade station j service water system consisting of three 50 percent capacity pumps. The ESW pumps take suction from the seismic Category I UHS retention pond. Locked closed isolation valves separate the two ESW trains.
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44- . Although each train of the ESW system is interconnected with the non- , safety-relatedstationservicewater(SSW) system,afteraSIS,'AFW low suction pressure, or loss of offsite power the ESW system is , automatically isolated from the SSW system.
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The ESW system heat loads are all_ safe.ty related exc.ept_for the_a.ir _.. - compressors and include the diesel generators, safety-related air 1 i conditioning systems, all safety-related pump room coolers, and the I containment air coolers. The nonseismic air compressors have seismic Category I isolation valvesseparating the essential and nonessential portions of the ESW system.
, U. 5, ACCEPTANCE CRITERIA
- 1. GDC2 " Design Bases for Protection Against Natural Phenomena" Both the CCW and ESW systems at SNUPPS are protected against natural phenomena. The systems are designed to seismic Category I requirements and are located in seismic Category I, wind and tornado missile pro-tected structures which also provide protection against the design basis flosi. All connections to nonessential portions of the systems are isolable by seismic Category I, automatically operated isolation valves such that a single failure following any natural phenomena will not prevent either system from perfonning its safety function. Regulatory Guides 1.26, " Quality Group Classification," 1.29, " Seismic Design Classification," 1.102, " Flood Protection for Nuclear Power Plants,"
and 1.117. " Tornado Design Classification," were followed in meeting i GDC2, e _.]
- 2. GDC4"EnvironmentalandMissileDesignBases[ ,
*Each CCW pump and ESW pump and all n'ecessary, piping and valving are located ih separate rooms or cubicles such that flooding,' environ-mental conditions, pipe whip and jet impingement due to a pipe break, or internally generated missiles could only affect components in that particular room or cubicle. Since both the CCW and ESW systems are classified as moderate energy, dual-purpose (used during nomal opera-tions and emergency conditions) systems, single active failures followingapipebreak(crack)inaseismicCategoryIportionofthe system are not postulated in accordance with ASB 3-1. Following a e
pipe break in a nonseismic Category I portion of the systems, no single active failure will result in the loss of more than one ESW or CCW train. Both systems are adequately separated from high energy piping systems such that pipe whip, jet impingement or missiles.. generated.-... . from high energy piping system are not a concern. I e 3. GDC5 " Sharing of Structures, Systems add Components" Therdis no sharing of the CCW system or ESW system between units, so this GDC is not applicable.
- 4. GDC44 " Cooling Wt?r For both the CCW $ystem ud the ESW system component redundancy is such that their safety function can be perfomed following a loss of offsite power coincident with any single active failure. Automatic isolation valves are provided to isolate nonessential portions of the I system from essential portions of the system. The combination of the CCW and ESW systems have adequate capacity to transfe- heat inads, including decay heat, from safety-related structurer ;ystem 4:1d components under both nomal opera' ting and accident ~conaji... .....
u l S. GDCs 44 and 45 " Inspection of Cooling. Water System".and " Testing of Cooling Water System" - The ESW and CCW systems are inspected and tested periodically in I accordance with plant Technical Specifications. The systems are capable of being tested through the entire sequence of operations that brings them into service including the sequencing of the components onto the emergency buses. Limiting conditions for operation are pro-vided that are consistent with the Westinghouse Standard Technical Specifications. The systems are designed and located in ac'essible c areas to allow periodic inservice inspection of the components and piping.
. Analysis - This topic is covered in a separate section.
. Areas for Further Exploration
- 1. Identify which specific safety concern led to the addition.:of special pumps in the CCW system to supplyonly the containment air coolers upon loss of all normal CCW pumps. Detemine whether consideration was also given, to providing this capability for other components cooled by the CCW system.
- 2. Ascertain whether the primary reason for adding the reserve ultimate heat sink was based on common mode ESW pump failure such as clogged screens, or whether it was based on the problems of designing the ESW system to seismic requirements.
I i
c Plant Auxiliary Electrical. Systems 7.C
Title:
Description:
The S'Izewell "B" vital ac electrical system consists of four essential
. .. buses, each o'f'which feeds one of the four em~ergency load groups. The four essential buses receive offsite or station power through the nonessential ttation buses. A normally-open bus tie between essential buses -1 & 4 and 2 & 3 is used to provide an alternate source of offsite power to an essential bus. The standby power sources to the four essential buses are four independent diesel generators, each of which is rated 5500 kw and supplies only one of the essential buses.
The Sizewell "B" vital de electrical system consists of eight independent batteries, each with its own battery charger, distribution boards and invert'er. Four of the batteries provide power through the inverters to the primary protection and instrumentation systen!. while the other four power the secondary protection and instrumentation system . The batteries also provide power to other essential equipment such as the steam driven auxiliary feedwater pumps, the diesel driven emergency charging pump, main and emergency control room lighting 3and control and switching power for essential equipment. Section 15.5.2.1 of the preconstruction safety report indicates the batteries have capacity to power the steam driven feedwater pumps and charging pump for 12 hours. The four de power trains which supply primary instrumentation and the four which supply secondary instrumentation are each electrically indepen-dent. The four ac power trains are electrically independent exceptt for the common source of offsite power to the buses and the crossties
. l between buses which are only enabled when on offsite power. The four ac ;
and de electrical trains are physically separated for fire protection purposes either four ways or two ways. Wherethereistbwayseparation, the equipment associated with trains 1 and 3 is separated from equip-ment on trains 2 and 4.' Within these pairs of safety trains it appears the physical separation generally follows that of IEEE 384. The primary and secondary reactor protection system batteries are located in separate buildings, however within the auxiliary building and reactor building, primary and secondary cables of the same channel will be run together. SNUPPS
Description:
The SNUPPS vital ac electrical system consists of two essential buses, each of which feeds one of the two completely redundant emergency load groups. The two essential buses receive both the nomal and alternate source of offsite power directly from the offsite power supplies. No bus tie exist between the essential buses. The standby power sources to the two essential buses are two independent diesel generators, each of which is rated 6200 km an'd supplies only one of the essential buses. The SNUPPS vital de electrical system consists of four independent batteries, each with its own battery charger, distribution boards and inverter. The batteries provide power through the inverters to the four channels of the reactor protection and engineered safety features systems. The batteries also provide power to other essential loads such as the steam driven auxiliary feedwater pump train, main control room
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. -19 emergency lighting and switching power for essential equ1pment. The batteries have sufficient capacity to power their toads for 3 hours and 20 minutes.
The four de power trains are electrically independent. The two ac power trains are electrically independent except for the common sources t of offsite power to 'the buses. The de and ac power trains maintain physical separation between redundant divisions in accordance with the
- criteria contained in IEEE 384.
U.S. Acceptance Criteria The U.S. acceptance criteria used by the staff for the vital onsite ac and de distribution systems is contained in Table 8-1 of the SRP and is primarily based on the General Design Criteria. The criteria requires that the onsite power systems have redundancy, meet the single failure criterion, be testable and have the capacity, capability and reliability to supply power to all the required safety loads. i The onsite ac power system at SNUPPS meets the redundancy and single l failure criterion by utilizing two independent,100% distribution system divisions, each of which is powered by an independent diesel
- l generator or by one of two offsite circuits. The diesel generators and offsite circuits meet the capacity and capability requirements by having sufficient capacity to sequentially start and operate all of the safety loads connected to then, The diesel generator in addition has demonstrated the required reliability and capability by meeting the requirements of IEEE 387 and RG 1.9 for automatic sequential loading, load rejection, light loading, reliability qualification
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testing, margin test, and load capability qualification tests. The onsite de power system at SHUPPS meets the redundancy and sing'le failure criterion by utiliziry four independent distribution system divisions, each of which is powered by an independent battery and battery charger. Each division meets the capacity and capability requirements by utilizing battery chargers sized with sufficient capacity to supply the largest combined demand of all the steady state loads connected to it. Although no specific criteria exists for battery endurance (following completion of Station Blackout USI A-44, specific crite,ria will be developed) it is required that the turbine-
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driven AIN pump t' rain.bE aiailable for'2 hours assumino a total. loss.cf ac power. The SNUPPS batteries, which support the turbine driven AFV
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pump train, have a'3 hour and 20'ininntes entrance whicimeets this requirement. The redundancy and reliability requirements are further met by the ac and de vital onsite systems at SNUPPS by physically separating the - power supplies, to the redundant divisin'ns in, senarata rams in a seismic.. - Category I building and physically separating switchgear and cabling of redundant divisions according to IEEE 384 and RG 1.75 requirements. 1 i Analysis Provided separately i l l L s
Area for Further Exploration (1) It is not clear whether those cable runs which have only two wayseparationforfiresegregationpurposesstilli$ fact - maintain four way separation for electrical segregation purposes as defined in Section 8.4.2.3.(b) of the Pre-Construction Report. i Determine whether there is four way electrical segregation in these areas and if this results in four way electrical segregation for all the safety equipment. (2) Chapter 15 of the Pre-Construction Report indicates that the essential batteries have capacity to power the steam driven AFW pumps and charging pump for 12 hours, and Chapter 8 of the report indicates that the battery rating has not been finally
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established. Determine whether the 12 hour endurance quoted for the batteries which support the AFW pump trains include all other safety loads connected to the batteries and whether the batteries not associated with the AFW and charging pumps also have 12 hours , endurance. I-6 5 4 4
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JAll 2 * .T33 MEMORANDUM FOR: B. J. Youngblood, Chief ' Licensing Branch No. 1 Division of Licensing FROM: G. E. Edison, Project Manager Licensing Branch No. 1 - Division of Licensing
SUBJECT:
SUMMARY
OF MEETING WITH WESTINGHOUSE ON REACTOR COOLANT PUMP SEAL PERFORMANCE (GENERIC) The subject meeting was held in Bethesda on January' 19, 1983. The purpose of the meeting was to gather information on the performance of Westinghouse Reactor Coolant Pump (RCP) Seals. This information was needed to assist in the evaluation of the probability of certain accident scenarios involving RCP seal leakage. Westin' house made an impressive presentation in which they described the RCP and seals, the seal support functions and systems, seal performances and operating experience, and systems reliability and accident scenario probabilities. They indicated that the RCP was designed to operate without damage to the seals for at least 24 hours, and probably much longer, if seal injection water was lost while-component cooling water to the seal thermal barriers was maintained, or vice-versa. Test results were presented for valve seal "0" rings which showed
- essentially no damage (slight feathering) after 10 hours at 550 F; the "0" rings were stated to be identical in material, thickness and include the range of clear-ances that exist in RCP applications. The tests were stated by Westinghouse to be extrapolable to RCP application. Further RCP seal tests and analyses are planned.
Using seal failure rates based on the valve seal "0" ring test data as well as reactor operating data, Westinghouse presented analyses which were stated to conservatively predict a 10 gpm leak rate for as long as 10 hours if all cooling water to the RCP, seals (component cooling water to seal thermal barriers, plus CVCS seal injection flow) were shut off. The analyses assumed the RCP was tripped and the seal "0" rings did not fail. Westinghodse stated that they had done a bounding analysis earlier for Indian Point in which they assumed that all three seals were totally missing from the RCP; for that case they calculated a maximum leak rate of 300 gpm per RCP. They . stated this was the largest leak possible which could result from a loss of all seal cooling water. They noted that an earlier Westinghouse calculation for the large leak at the H. B. Robinson plant had estimated a leak rate of about 400 gpm; however, they indicated they had little faith in that calculation, that they could not find it now, and the individual who had performe4 it is no longer with Westinghouse. The Robinson event did not involve a loss of all seal cooling water. - 4
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3L._..T..i__..______..._- B. J. Youngblood hut off' all seal cooling water, a For the loss of all AC power (which would g/R-yr. was estimateg, nationwide average probability of S2 X 10- . res probability of core uncovery (after RCP seal LOCA) of s3 X 10 /R-yr. The analy-ses used EPRI data for recovery of AC power, and showed the loss of all AC p,ower to be the strongly dominant mode of losing all seal cooling water. Pipe break was not considered to cause a loss of all seal cooling water; it was noted that in the SNUPPS design, no single pipe break in the CCW system would cause a loss of CCW function because of the redundant flow trains. It was also noted that in the SNUPPS design, d.c. battery endurance was 200 minutes fully loaded, and 8 to 12 hours with load shedding; also, the turbine-driven auxiliary feedwater train could be manually operated without batteries.
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Westinghouse concluded that: (1) The design of the RCP seal system and plant seal support systems meet the licensing design basis criteria. (2) Available data and operating experience show that the probability of signif-icant seal leakage (from loss of seal cooling water) is low. (3) Loss of ~ all a.c. power is judged to be the dominant cause of failure of all seal cooling. (4) Th'e probability of core uncgvery after loss of all a.c. power (due to RCP seal leakage) is N3.5 X 10 /R-yr. The Westin,ghouse presentation is attached, as well as a meeting agenda and a list of attendees. G. E. Edison, Project Manager
. Licensing Branch No.1 Division of Licensing Attachments:
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8 ATTENDANCE AT WESTINGHOUSE /NRC MEETING HELD JANUARY 19, 1983
,0N REACTOR COOLANT PUMP SEAL PERFORMANCE NRC Particioants Kansas Gas & Electric G. Edison,' DL G. Rathbun T. Speis, DST .
M. Stewart E. Goodwin, NRR J. Bailey A. Thadani, DST . R. Riggs, DST Kansas City Power & L-ight Co. L. Marsh, DSI
- A. Busiik, DST F. Crawford A. Singh, DSI .
B. LeFave, DSI -Bechtel J. Wermeil, DSI R. Anand, DSI P. Ward C. Liang, DSI J. Prebula J. Holonich, DL D. Quattrociocchi Westinghouse SNUPPS D. Rawlins . F. Schwoerer v
- W. Poulson R. Etling D..Salak D. .Paddleford J. Crane *
' W. Brown G. Harkness ,
J. Swogger Uni'on Electric A. Passwater S. Miltenberger D. Capone i A g e e. 9
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- 1. DESIGN OF RCP SEAL SYSTS AND PLANT SEAL SUPPORT SYSTEMS EET THE LICENSING BASIS DESIGN CRITERIA,
- 2. EXPERIENCE AND AVAILABLE DATA SHOW EAT THE PROBABILITY OF SIGNIFICANT SEAL LEAKAGE IS LOW, ( CURRENT STUDIES ARE EXPECTED TOCONFIRMTHISCONCLUSION.)
- 3. THE OCCURENE OF LOSS-OF-ALL-AC POWER IS JUDGED TO BE THE DWINANT CONTRIBUTOR LEADING TO FAILURE OF ALL SEAL COOLING,
- 4. THE CALCULATED FREQJENCY'0F CORE UNC0VERY FRm THE LOSS-OF-AC-
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, ACCIDENT DUE TO RCP SEAL LEAKAGE IS APPROXIPATELY 3,5 x 10 /YR.
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AGEllDA Westinghouse /flRC Meeting January 19, 1983, 9AM P110 Phillips Bldg Reactor Coolant Pump Seal Performance Approximate Presentor Presentation Time
- 1. Introduction W NTD 5 Min.
- 2. RCP - Function and Features W EMD 20 Min.
- a. Pump
- b. Seals
- 3. Seal Support Functions W NTD 15 Min.
- a. Seal Injection j b. Thermal Barrier -
- 4. Seal Performance and Experience W EMD 60 Min. _
! a. Operating History l
- b. O Ring Test Data -
- c. Projected Seal Performance l During Upset Modes i
! d. WOG Test and Analysis S. Seal Support Systems ]!,NTD/SNUPPS 20 Min, j- a. CVCS
- b. CCW l
'6. Systems Reliability W NTD/SNUPPS 40 Min.
l' a. Scenarios Including Full Loss of AC Power -
. b. Support Systems
- c. Seal ..,..,
- k d. NSSS Effects -
3.a
- e. Recovery / Mitigation Actions f^@
> 7. Conclusions 10 Min.
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r .r> REACTOR COOLANT PUMP GENERAL DESCRIPTION THE CONTROLLED LEAKAGE SEAL REACTOR COOLANT PUMP IS A VERTICAL, SINGLE-STAGE, CENTRIFUGAL PUMP DESIGNED TO MOVE LARGE VOLUMES OF REACTOR COOLANT WATER AT ELEVATED TEMPERATURES AND PRESSURES.
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TYPICAL P 2200 PSI TYPICAL TEMPERATURs 155'F
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i < t REACTOR COOLANT PUMP SEAL SUPPORT FUNCTIONS WESTI'GHOUSE N JANUARY 1983 .
REACTOR COOLANT PUMP SEAL SUPPORT O REQUIRED TO MAINTAIN RCP SEALS COOL AND CLEAN
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ENSUR~S SEAL FACES MAINTAIN PROPER SEPARATION PROVIDES A CLOSED LOOP PROCESSING FOR ALL PUMP LEAK 0FF'S
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-0 SUPPLY CHARACTERISTICS ,
0 0 TEMPERATURE BETWEEN 60 F AND 130 F l FLOWRATEBETWEEN6GPMhD13GPM
, PRESSURE SUFFICIENT TO ENSUR'E REQUIRED FLOWRATES (GREATER THAN RCS PRESSURE) 0 PROVIDED BY THE CHEMICAL AND VOLUME CONTROL SYSTEM CHARGING PUMPS t
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s RCP THERMAL BARRIER COOLING O REQUIRED DURING NORMAL RCP OPERATION O PROVIDES REDUNDANT SEAL COOLING WHEN REACTOR COOLANT 0 TEMPERATURE IS ABOVE 150 F 0 LIMITS THE HEAT TRANSFER FROM THE REACTOR COOLANT TO THE LOWER PUMP INTERNAL COMPONENTS 0 SUPPLY CHARACTERISTICS , 0 COOLING ~ WATER TEMPERATURE BETWEEN 80 F AND 105cF NOMINAL FLOWRATE OF 40 GPM LOW PRESSURE SOURCE O THERMAL BARRIER COOLING IS PROVIDED BY THE COMPONENT COOLING WATER SYSTEM
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CHEMICAL AND VOLUME CONTROL SYSTEM m e en. 4 es e e. i . WESTINGHOUSE JANUARY 1983 9 9 4 6 o
CHEMICAL AND VOLUME CONTROL SYSTEMS FUNCTIONS o WESTINGHOUSE NSSS DESIGN PROVIDES THE FOLLOWING FUNCTIONS: MAINTAINS REACTOR COOLANT INVENTORY t
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REACTIVITY CONTROL - PORTIONS USED FOR SAFETY INJECTION t REACTOR C00LANT PUMP SEAL SUPPORT e e **..
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SUMMARY
OF CHEMICAL AND VOLUME . . CONTROL SYSTEMS FEATURES O REDUNDANCY 0 SAFETY CLASS 2 0 SEISMICALLY QUALIFIED ACTIVE PUMPS 0 EMERGENCY POWER t 4 ' o . 9 1 e e m
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_ . .l . . _ .- _. . . _ ,_ _ . COMFONENT COOLING WATER SYSTEM _ FUNCTIONAL REQUIREMENTS 4 _ m O 0 as O O + WESTINGHOUSE JANUARY 1983
COMPONENT'C00 LING WATER O A SYSTEM PROVIDING A CONTINUOUS SUPPLY OF COOLING WATER TO PLANT COMPONENTS HANDLING POTENTIALLY. RADIOACTIVE FLUIDS 0 FORMS AN INTERMEDIATE BARRIER SETWEEN THE ULTIMATE HEAT SINK AND POTENTIALLY RADI0ACTI'/E SYSTEMS i e e em .h >i l
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- W ' h' % edw'. *a*.. .N w p ag e gJ+ em4 ad e J e*, - a a * * - * * * * * * *** * *ee*= *W- wer' e we eequemeesw w'.@ h eta,* gm - ^*- '*
3 e
4 FUNCTIONS 0F COMPONENT ' COOLING WATER SYSTEMS f
- 1. REMOVAL OF RESIDUAL AND SENSIBLE HEAT FROM THE REACTOR COOLANT SYSTEM THROUGH THE RESIDUAL
~
HEAT SYSTEM -
- 2. COOLING OF SAFEGUARDS EQUIPMENT AND HEAT LOADS
. FOLLOWING DESIGN BASES EVENTS. -
r
- 3. HEAT REMOVAL FROM VARIOUS NSSS AND BOP COMPONENTS DURING NORMAL PLANT OPERATIONS O
t i.
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INDUSTRY DESIGN CRITERIA / RECOMMENDATIONS COMPONENT COOLING WATER SYSTEMS
- 1. COMPONENT COOLING WATER IS NUCLEAR SAFETY RELATED
- 2. SAFETY RELATED COMPONENTS MUST HAVE SEISMIC SUPPORT
- 3. SYSTEMS MUST INCLUDE REDUNDANCY FOR SINGLE FAILURE PROTECTION
- 4. CCW PUMPS MUST HAVE AUTOMATIC ACCESS TO EMERGENCY .
POWER SUPPLYS
- 5. A SAFETY RELATED MAKE-UP SOURCE (1UST BE AVAILABLE
, FOR PASSIVE FAILURE LEAKAGES
- 6. SUBSYSTEMS WITH PHYSICAL SEPARATION FOR RECIRCULATION
, SHOULD BE' INCLUDED o
l
- 7. SUFFICIENT REDUNDANCY SHOULD BE INSTALLED FOR ON-LINE -
MAINTENANCE CAPABILITY l :
_ _.s- . . _ _ _ _. . . . . . . . - _ - . __ - _ . . _=. - - - - , l NSSS COMPONENTS TYPICALLY - SERVICED BY COMPONENT COOLING
- 1. REACTOR COOLANT PUMPS
- 2. RESIDUAL HEAT EXCHANGERS 3< EMERGENCY CORE COOLING SYSTEM PUMPS
- 4. CHEMICAL AND VOLUME CONTROL SYSTEM COMPONENTS
- 5. WASTE DISPOSAL COMPONENTS 6~. SPENT FUEL COOLING SYSTEMS 4
- 7. CONTAINMENT FAN COOLERS
- 8. AUXILIARY EQUIPMENT 9
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_ _ _ _ _ _ _ _ , ,p
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RCP'S
, C0tlTAINMENL - l [h_
' ~ ~ ~ ~ ~ ~ - - ~ ~ "
" SERVICE
-_hlATE o
~
ESF TRAIN COMPONENT COOLING C0f1PONENT COOLING WATER PUMP WATER llEAT EXCilANGER
'S 'S m _.hj. , NON- ___hIq_ u ,
l ESSENTJALS_ - 1 TYPICAL C0fiPONElll~ COOLING l-lATER SUBSYSTEti ; i; i i
,..... i
.. .. . ~ . . . . . . - - - . - - - - - - - . . - - - - - . . ..--.. .. .
RESPONSIBILITY FOR DESIGN 0 COMPONENT COOLING WATER SYSTEMS ARE INCLUDED IN THE OPTIONAL SCOPE FOR WESTINGHOUSE NSSS DESIGNS O DESIGN CRITERIA AND GUIDELINES ARE PROVIDED TO THE CUSTOMER /AE WHEN THE COMP 0tiiNT COOLING WATER DESIGil IS NOT WITHIN THE' WESTINGHOUSE SCOPE OF SUPPLY N -
*%.M A.
\
t s< e .y % e 1, ..__. .... _
.e f_ +
w a c_
GENERAL INTERFACE REQUIREMENTS
- 1. COMPONENT COOLING WATER FUNCTIONS
- 2. COMPONENT DESIGN REQUIREMENTS
- 3. SYSTEM DESIGN REQUIREMENTS
- 4. WATER CHEMISTRY CRITERIA
- 5. INSTRUMENTATION AND CONTROL,REQUIRENENTS
- 6. EMERGENCY ELECTRICAL POWER SUPPLY
. 7. HEAT LOAD REQUIREMENTS g, ,
sp e 4 e
~
COMPONENT COOLING WATER O SYSTEM PROVIDES COOLING TO NSSS AND BOP COMPONENTS DURING ALL MODES OF OPERATION 0 SYSTEM DESIGN COMPLIES WITH ALL CRITERIA REQUIRED OF ENGINEERED SAFETY FEATURES O RESPONSIBILITY FOR DESIGN HAS TRADITIONALLY BEEN
= PROVIDED BY BOTH WESTINGHOUSE AND THE CUSTOMER /AE O GENERAL INTERFACE REQUIREMENTS ARE PROVIDED WHEN DESIGN IS WITHIN THE CUSTOMER /AE SCOPE w
S W m i ,
mg he--a -e e d w+ - s-+-a sa** s e 6 9 e 6 e
- 4. SEAL PERFORMANCE AND EXPERIENCE
=
WEMD D. W. SAL:AK T
.. a ;
%4 e'*, "W e # 8
- 1 . . .
t e e s
- . ~
= .
~
CONTROLLED LEAKAGE SEAL REACTOR COOLANT PUMP ACCUMULATED EXPERIENCE e Over 5 million operating hours e Over 131 pumps in commercial operation , - e Over 43 plants in commercial operation with Westinghouse reactor coolant pumps
- First commercial ~ operation in 1968
- O 6
i.
;j . .
L
. _ . . .=_. _.__. _
- 7. _ ..
; . ~ . '.* .
REACTOR COOLANT PUMP MODELS WESTINGHOUSE REACTOR COOLANT SHAFT SEAL PUMPS
~
CURRENTLYINOPERATI0NORbNORDER APPROXIMATE FLOW INITIAL MODEL MODEL NUMBER NUMBER OF UNITS (GPM) STARTUP DATE 63 9 63,000 19.67 70 5 70,000 1968 93 23 90,000 1969 93A 150 100,000 _3.97.1.__ _ 93D 27 , 95,000 1974 93A-1 62 100,000 .H8C 100 17 106,000 1980. . .. TOTAL 293
.m.
e,_
1
.~ .
l i
. \
WMT
. LOSS OF A/C POWER .
9
. SEAL INJECTION AND CCW LOST
. RCP STATIC
. RCS AT FULL TEMPERATURE AND PRESSURE f
a . ji rl f *' . 4'
++e
' NS eh4 F M e e88 e 4 e 4 pDeoV b' tateme D '*O. + e se a% m ,gg, ,,
t.
, . ,, , , n , .,.. ---., - - - ., - " - - , , ,_ .,,n-.. - -
, n.._ . .. - -. - . . - , _ _ . . . . .-
t .- LOSS OF SEAL SUPPORT HISTORY TIME PUMP EVENT DATE (MIN.) PUMPS (MIN.) A 4/27/68 10 4 40 B 5/17/69 30 2 60 C 7/15/69 2 4 8 D 1/28/71 65 2 130 D 1/28/71 45- 2 90 E 3/14/71 SEVERAL 3 9 F 9/1/77 8 4 32 G 1/12/82. >9 3 27 24 396
=
I e m O N h
. .__ ._ .. . _ _ . . _ . _ _ __.m.-...
' LOSS OF SEAL SUPPORT HISTORY
SUMMARY
. 3 INSTANCES WITH CORE LOADED
. TOTAL OF 57 PUMP-MINUTES WITH CORE LOADED
~
. NO INSTANCES AFTER 1971 WITH CORE LOADED
. 4 PUMPS WERE OPERATED AFTER INCIDENT
. NO. 1 SEALS REUSED AT 4 SITES (OTHER SITES, NO RECORD)
. NO. 2 SEALS ALL REUSED AT 4 SITES EXCEPT
~
WARPED RUNNER ON ONE PUMP, HEAT CHECKED RUNNER ON ONE PUMP (OTHER SITES, NO RECORD) *
. NO. 3 SEALS HAD NO DAMAGE REPORTED iii s
9 e t f
7 , 0-RING EXPERIENCE DURING VALVE TESTS
. SEAL 29-INCH END PLUGS 1
1
. 4 TESTS (2 END PLUGS ON 2 VALVES) i 1
. 10 HOURS /550*F/2250 PSI) -
. SAME MATERIAL (80 IRHD ETHYLENE PROPYLENE)
. EXTRUSION GAPS LARGER THAN THOSE FOR SEALS NO 0-RING BLOWOUTS
, . No MAJOR EXTRUSION (FsATHERED EDGE ONLY) f ,
b
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. . _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ __-..___.__...A '
UPSET MODE FACTORS AFFECTING SEAL LEAKAGE FACTOR EFFECT ON LEAKAGE COMMENT VISCOSITY INCREASE RELATED TO SEAL WATER TEMPERATURE
. ~ TURBULENCE ." INCREASE. RELATED TO SEAL WATER TEMPERATURE THERMAL GRADIENT INCREASE SHORT-TERM CONDITION Two-PHASE FL0w DECREASE OCCURS ABOVE 365'F-BETwEEN.FACEPLATES 6664 6
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r - -- - - - - - - - . _ . . _ _ . _ ,_ _, , ,,
. i RC_P SEAL C0flSIDERATIONS 0-RINGS
. EXTRUSION
. MATERIAL PROPERTIES AT HIGH TEMPERATURE FLUID FLOW AT HIGH TEMPERATURE
. VISCOSITY
. REYNOLDS' NUMBER THERMAL GRADIENTS
- . FACEPLATE TAPERS *
. CLEARANCES / INTERFERENCE
. AXIAL SHAFT EXPANSION
. GRAPHITAR SHRINK FITS
. , CRUD EFFECTS
. BLOCKNo.1SdAS
. SEAL RING AXIAL MOTION.a. , -
!L ~ ~~ - - - ~, s .. .....w.
~ -. _ , *~~f.**__
. -~. -cA - . . #1 LEAK .0F.F LINS. PRESSURE ~. _ . . . . . _ . . . - . _ _ .
~
g
. _ RELIEF VALVE OPERATION AT 150 PSIG i-
. 2-hHASEFLOWATNo.2 SEAL
1 l HIGHPRESSURE-liIGHTEMPERATURE
. PARAMETRIC SURVIVAL TEST
, 2250 PSIG a 550 F
. 10-HOUR. DURATION
. ETHYLENE PROPYLENE SEALS
. IEFLON SEALS EXTRUSION GAPS WITHIN DES _!GN RANGE
. EXTRUSION GAPS EXCEEDING DESIGN RANGE _
. QUANTIFY DEGREE'0F PARTIAL EXTRUSION
. DETERMINE CRITICAL EXTRUSION GAP FOR LOSS OF SEAL ,
. MEASURE BEFORE AND AFTER ELASTOMER. HARDNESS
. PERFORM ON FULL SCALE TEST FACILITY a
s a e c 5 \
~ ' ~~
SUMMARY
. WITH OVER 300 MILLION PUMP-MINUTES OF OPERATING EXPERIENCE, ONLY 396 PUMP-MINUTES WERE ACCUMULATED WITH LOSS OF BOTH SEAL INJECTION AND CCW.
. 339 PUMP-MINUTES BEFORE CORE WAS LOADED
. LONGEST EVENT WAS 65 MINUTES
, NO DAMAGE TO NO. 1 SEALS
. 0-RING SURVIVAL FOR 10 HOURS IN VALVE TESTS
. EXPECTED SEAL LEAKRATE DURING LOSS OF BOTH SEAL INJECTION AND CCW IS LESS THAN 20 GPM/RCP.
. BASED ON 0-RING FUNCTIONAL SURVIVAL
~
. BASED.ON EXPERIENCE
. BASED ON ANALYSIS
. HIGH PRESSURE-HIGH TEMPERATURE PARAMETRIC SURVIVAL TEST WILL DETERMINE 0-RING FUNCTIONAL SURVIVABILITY FOR VARIOUS
, DURATIONS AND EXTRUSION GAPS.
. CONFI_RMATORY TEST
. EXTRUSION GAPS.BEYOND DESIGN RANGE
' " ~
. COMPARISON OF ELASTOMER PROPERTIES BEFORE AND AFTER TESTS .
. QUANTIFICATION OF EXTRUSION AND RELATED LEAKRATE
}_
- 3
\ + _ _ . _ __ ___.. ___7.__.
-.a . , . _ m __
--d 6 4 6- , w s. We9 m n e. i m Msmes P. s , eg y.. g y , g ,., ,
" e oe
% y 9
I* m O SYSTEMS REUABILITY 4 SEalBCES IBDING TO LOSS OF SEAL COOLING SEAL LEAK PROBEILITIES WITH TIME 6 I PROBABILITY T SEAL IfAKS #0 CORE ifC0VERY IN LOSS OF All AC SEW 9EES p 0 e
- r. .; - ,~. =.
.veS' emod W om.W o w' A -s-emommme so p ate)., eas ss es. ame * **M =e - ' s * * *O '
'-%s eo s- *'d . serm.me m ee ..a.W w easm=o -d'o ~ *^** * '* j'y ,
~
O
. . , .___ . _ . , . .-, ,, _. ._ _. _ m.y. ., ., y_ , _ . -- , . . -
1 Fall 2E OF SEAL
; COOLING L
1 O ~ 2.6x10-4 MAR
. -s l REACTOR TRIP TRANSIENT LDSS T WFSITE F0ER INITIATOR IS [ DSS T SRRIOUS CONTAltifNT
, INITIATOR G0 MAR) INITIATOR (0.12 MAR) OPERATING Col RfP ISOLATION SIGML (0.12 MAR) INITIATOR (.01 MAR) 0 2.53 x 104M AR ,
< 4 x 104MAR
~
1 x 10-7 MAR ~s 2 x 10-6 MAR I mm um v2 m START START START - 2 x 104/YR S x 10-5/YR 3 x 10-6fyg 98% T RCP SEAL COOLING FAllME IS ATTRIIUTEDTOLOSPINITIATOR b /
. , . . . i .
- ~ -
~
_7 7 . .. .._ _ .._
' m EXPERIENCE WITH 80 DUlDtitM 0-RIfES AT P Af0 T ON RCP DURIPE Pl#ff ItCIDENTS 8 IfCIDENTS VARYING FRCN 2 MllUTES TO 65 MItUTES WITH0lfT SFA C00 LITE - GI'E 396 REP MIN. (6.6 RfP HRS.)
- GIVE E6 0-RIFE HRS'. SITE 10 0-RITES /RfP f%INLOOP STOP VAltE TESTS 2 MAINLOOP STOP VALVES EACH WITH 2.SFAS WITH 3 0-RIBES EACH WERE lt ilti) AT 2250 PSIG, 550*F FOR 10 HOURS GIVES 170 0-RING HOURS.
ALTHOUGH THESE 0-RITES HAVE ~ 3 x DI#EIER (PERIMETER) 0F RCP 0-RINGS, SITE 3 0-RINGS ACTED IN SERIES, THIS IS CONSIDERED A TPADEDFF. G A i, . .... _.... O O q
~
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s f 04IPE FAIL 11PE RATE AT P AfD T 186 0-RItG HOURS WITHOUT C00 LITE - NO FAlli!RES 1 50 = 3.74 x 10-3/0-RIfG HR.
^95 = 1.6 x 10-2 /0 -RIBG HR._
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SIGNIFICANT SF.AL EAK PATliS
-EE HOUSIfG y
- 0-Rile C
RUfER 20 - 100 GR Cl.MP-RItG c / 100 - 300 GM C / 20 - 100 G M SEM. RIFE CLMP-RIPE N[ C w. c / DB_TA
- lEL P = xT = PROB. OF AN 0-RIfG FAILIfG IN TIE T
[fAK RATE FOR PROB ONE RCP - EXPRESSIM 1 HR 3 HR 6 HR 10 HR OGM 1-6P .978 .93t4 .868 J/8 100 SP 1.85 x 10-2 5.5x10-2 1.1x10-1 1.85x10-1 200 8P2 NEGLIGIBE 300 P 3.7x10-3 1.1x10-2 2.2 x 10-2 3.7x10-2, E P2 fELIGIBE _ L i fr
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. FWR RfP LEAKS LEAKRATE EBOL. 12 3 HR 6 HR 10 HR O 1 - 25P .9085 ._725 A5 .075 100 20P 7.4x10-2 2.2x10-1 4 A x 10-1 7.4x10-1 300 4P 1A8 x 10-2 4A x 10-2 8.8x10-2 1A8 x 10-1 1200 P* 3.7x10-3 1.1x10-2 2.2x10-2 3.7x10-2
- ASSLED COPLETE COUPUNS ACROSS PUPS FOR 1200 GFN CASE.
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- P (100 gpml -
+ 9 P (300 gpm) 10~I - e a200 ,,.> y t . .I 2' - 10-2 , e s, .M rf e
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i :1 10-3 . . . . . . . . . . . . . . . . . 3 6 10 18 1
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N 9 # e e- % 9 FREWENCY = 0.122/ YEAR (EPRI NP-230D TIE AT B0 PROB. RESTORED PROB.NOTRESTORED OF INTERVAL IN INTERVAL BY T
.1/2 HR. .48 .52 1 .14 .38 2 .1 .28 4 .05 .23 8 .14 .09
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. DIFP GEfBATOR Fall 1RE AlWOW-SE/ERSON MODEL NUPEG/CR-2099 P = (xD2+RT 2 A
R
= 7.5 x 10-5/HR.
2 = 2 x 10-6/HR.
]j P = 1/3(AD 2+ 1/2R2T T P 180 HOURS .
2.4 x 104 360 HOURS 6.03x104
. 720 HOURS l1.7x10-3 . DIESEL GEf6ATOR REPAIR EPRINP-2433 , TIE PR REPAIRED IN INTERVAL PR NOT REPAIRED BY TMAX
~' 0 - 1 HOUR .23 .77 1-4 .11 .66 4-8 .23 .43 8 - 12 .13 .30 3 , 12 - 16 .: .02 .28 16-20..; .08 .20
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P Diesel Not Repaired 10'I - P Offsite Not Restored E p
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E E Q - Ei* P Offsite Nor Specific y Diesel Restored b
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P Neither of 2 DG Nor i Offsite Restored i .- i f. 10-3 , , , , , , , , . , . u z 4 6- 8 10 12 14 16 18 20 22 24 26 28 Time (hours) o
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- 6 OWrE rF WAK TYPE T
PL- (x -ATd T) (P (T)) o 24
- A Pg(T) dT = 2.32A 0
4 0 m g WHEE P (T) BASED ON OFFSITE POWER OR SIfEE DG RESTOPATION 4 Pm
- - = 0,17 Pg = 0.054 Pm = 0.0085' i
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] 200 300 RCP Seal Leakage (gpm/ pump) '-
!. Estima'ted Time to Core Uncovery Following Loss of All AC Power t
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=
- CHANCECOREUNC0VERYAS RESULT CF SEAL EAKS lfAK T = TIME TO P (UNC0VERY)
PLR UNCD/ER CORE TOP Ppp(T) =P LR X Pop PEE _ 60 GRi .788 18.5 1.25x10-2 9.8x10-3 100 .17 14.0 2.3x10-2 3.9x10-3 300 .034 7.0 5.2x10-2 1.8x10-3 1200 .0085 2.0 .21 1.8x10-3 TOTAL 1.7x10-2 FREO. CORE UNC0VERY = F(ALL AC) X Pg g y 4
= 2.1 x 104 x 1.7 x 10-2 = 3.5 x 10 /YR.
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INITIATIPE EVENTS IIADIWrTO A-LOSS
& SEAL C00 LITE
- 1. LOSS OF OPEPATIfG Cr?!POENT C00LItG WATER RN FREW ENCY = 1.2 x 10-1/ YEAR
- 2. SRJRIOUS HIGH-HIGH CONTAIWBff PRESSURE SIGML FREWEfCY = 5.0 x 10-3/ YEAR
- 3. NON-LOSS OF 0FFSITE POWER TRANSIENTS 1
FREWENCY = 1.0'x 10 / YEAR
- 4. LOSS OF 0FFSITE AC POWER, FWR POWER STATES ANALYZED:
FREWBCY = 1.2 x 10-1/ YEAR A. LOSS & ALL AC POWER, FREWENCY = 2.1 x 10 4 / YEAR B. LOSS OF OE (EITER) TRAIN DIESEL GEERATOR, FREWENCY =
.- 8.4x10-3/ YEAR C. LDSS OF EITER DIESEL GEEFATOR SET, FREWE?CY = 1.1 x 10-1/ YEAR E
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LOSS (FOPEPATItE COPnfBIT C00 LIFE WATER RNP PACK-UP AlITO-START PWP IN SNE TRAIN: PROPABILIT/ W FAILURE TO START = 3 X 10-3 PROPABILITY W PLNP IN MAINTENANCE = 2 X 10-3 OPERATOR CAN ALIGN REDUNDANT CCW TPAIN: PROPABILITY OF OPEPATOR FAILURE = 1 X 10-3 SIDEE VALVE FAULTS = 2 x 10-3 INDICATIONS AVAILAELE: .
- FLOW AIARM5, CCW
- PRESSURE ALARMS, CCW
- POSITION LIGHTS, C04
- TEMPEPATURE ALARMS, REACTOR COOLANT PLNPS
- CmPUTER MONITORED INTERLOCKS s .~ ~
' FREDlBEY
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T LOSS OF SEAL C00 LIFE IS 2 x 104 m a = esm.
-~----~~------~-:
x-- SRJRIWS CIS-B SIC #1 CBURIRJGAL CHARGING RNPS START, AVAILABLE IF POSITIVE DISR.ACEMENT R.PP FAILS, NO INTERRllPTION OF SEAL IRJECTION FLOW, (4.5 x 104/D8 W 0) OPERATOR 1%S CAPABILITY TO BYPASS CIS-B AND ALIGN CCW TO THEFMAL BARRIER COOLERS.
- 4 FREQJENCY OF LOSS T SEAL COOLING IS 4,5 x 10 / YEAR. -
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4 . . NON-LOSS OF 0FFSITE FG8 TPANSIENTS BOTH C0HEIT COOLING WATER AND 0%RGING SYSTEMS OPEPATE IN NOR%L MODE. ND EXCESSIVE D8WES ON SYSTEMS. FREQJENCY OF LOSS T SEAL COOLING IS 1.2 x 10-7 OPEFATING CCW RN MU5T FAIL 1.2x104 BA011P CCW PTP FAIL TO AUTO-START 6x10-3 ALTERNE TRAIN lilt STARTED BY OPERATOR 3.1x104
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l . l' LOSS T TFSITE AC P0le i l f l LOSS OF ALL AC POWER - SYSTEMS UtRVAILABILITY DETERilED BY POWER ECOVERY LOSS T OE TPAIN DIESEL GEEFATOR - LOSS T SEAL COnLItG D011 FATED l BY FAlllJE OF ESSENTIAL SERVICE FAER PUMP, OR FAILURE OF P0l FRED 1 l CTP0ENT C00 LIFE FATER PLIP TPAIN PROEABILITYOFFAILURETOSTART = 3x10-3 PROBABILITY OF PLPF(S) IN MAIN 1BRNCE = 2.0 x 10-3 LOSS OF EITER DIESEL GEEPATOR - ALL SYSTEMS A9AILABLE, CCW RJ1PS l AUT0-START ON LOOP, - i G
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