ML20028G346
| ML20028G346 | |
| Person / Time | |
|---|---|
| Site: | Clinch River |
| Issue date: | 09/17/1982 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| 20183, ACRS-2033, NUDOCS 8302080465 | |
| Download: ML20028G346 (31) | |
Text
{{#Wiki_filter:_ A?dRS ROSS PM6$683 ., m l L s- ) / i c O d Date Issued: 9/17/82 MEETING MINUTES FOR THE CRBR WORKING GROUP ON SYSTEMS INTEGRATION AND INSTRUMENTATION CONTROL SEPTEMBER 30, 1982 - WASHINGTON, D.C. The ACRS CRBR Working Group on Systems Integration and Instrumentation and Control held a meeting on September 30, 1982 in Room 1167, 1717-H Street, N.W., Washington, D.C. Working Group was briefed by the CRBR Project Staff and the NRC Staff on the proposed corfiguration of plant protection and instrument control systems for the CRBR. This was intended to be the first of a series of meetings during which the Working Group would review the CRBR Plant Protection and Instrumentation and Control Systems. The meeting was attended by W. Kerr, Chainnan, M. Carbon, J. Ray, J. Ebersole, and D. Ward, Working Group Mem ers; R. Savio, P. Boehnert. ACRS Staff; A. Bice, ACRS Fellow, and W. Lipinski, ACRS Consultant. Notice of this meeting was published in the Federal Register on l September 13, 1982. A copy of the notice for this meeting is included as Attachment A and a list of.the attendees is included as Attachment B. The schedule for this meting is included as Attachment C. A complete set of handouts has been included in the ACRS files and select portions of the handouts are included as Attachment D. The meeting was begun at 1:00 p.m. with a short executive session in which Dr. Kerr the Working Group Chairman sunnarized the objectives for the meeting. The meeting was adjourned at 6:00 p.m. The meeting was conducted entirely in open 1 session. The Working Group heard presentations from the NRC - NRR Staff m m omm 8302080465 820917 hN \\ P l 2033 PDR etified By
L :. ~ Meeting Minutes on CRBR Working Group 12/10/82 ad fron. representatives of the Clinch River Breeder Reactor Plant 1 l Project. ~ Introduction, P. Lickson, Westinghouse /CRBR Project Mr. Dickson sumarized important characteristics of the CRBR plant protection and control systems. The CRBR control / safety rod systems have been designed to be faster acting than thos.e in the typical commercial LWR. As compared to a LWR, the CRBR uses coolant with a much lower specif#: heat, has a larger temperature drop across the core, and does not have the large negative temperature coefficients associated with a typical LWR. Pump coast-down in the CRBR n,ust be fairly rapid to l avoid thermal shock to the upper internal; structures. Two independent rod insertion syster!s are used to control reactively. A high degree of diversity has been incorporated into these systems. They use different rod types, different types of trip signals, and different actuating devices. Physical Features Relevant to CRBR Control and Protection System, R. Doncals Mr. Doncals dcscribed the physical configuration of the CRBR control and I protection systems. (The CRBR core layout is shown on figures 1 and 2 of Attachment D.) Nine primary control assemblies and six secondary control assemblies are utilized. The Primary Control Rod System (PCRS) is used 1) to shut the reactor down from hot full power to zero power at the hot shutdown temperature, 2) to compensate for excess reactivity loaded into the core for burnup and operational requirements, 3) to allow for the maximum reactivity fault associated with any anticipated +
p c Meeting Minutes on CRBR Working Group 3 12/10/82 occurrence (postulated to occur upon the accidental withdrawal of the highest korth control rei inserted in reactor, 4) to compensate for the ~ failure of any single active components (highest worth control rod stuck). The Secondary Control Rod System (SCRS) is used to: 1)shutthe reactor down from hot full power to zero power at the refueling temperature, 2) to compensate for the maximum reactivity faults, 3) to compensate fer the highest worth rod stuck. Qualitative reliability analysis has been performed for the control and ple.nt protection systems and has been used to improve the design of the-system. The use of numerical reliability goals has been abandoned. The design of the PCRS is similar to what is used in the FFTF design. The SCRS is a unique design. (7chematics of the PCRS and SCRS hardware are shown on the Figure 3 of Attachment D). Primary Control Rod System, G. Smith, W/CRBR Project Mr. Smith described the PCRS.. The functions of this system are to move control assemblies to control reactor power and to initiate a negative reactivity insertion upon the reccipt of a scram signal from the Plant Protection System. The primary control rods use a roller nut latch of the type shown on Figure 4 of Attachment D. They are inserted by gravity and spring forces. Tr.e primary control assembly is shown on Figure 5 of Attachnent D. The system is Seismic Category 1 and Safety Class 1 and hu, two independent position indicating systems. The seek-system which controls rod motion has selectable speeds between 0.36 and 9.0 inches per minute. Total withdrawal stroke will be between 36 and 37.8 inches. The design basis for the system stipulates that the drive mechanism be able to exert at least 1000 pounds of force to free a T v- .. ~
y. Meeting Minutes 01 CRBR Working Grow 12/10/82 i stuck rod. The PCRS is the primary shutdown system and will be capable of shutting the reactor down during an SSE. The response of the system will such that the required fuel damage limits are not exceeded independent of the action of the SCRS. Secondary Control Rod System, R. Lawrence, W/CRBR Project Mr. Lawrence described the SCRS system. The system is intended as a second fast acting shetdown system, and is intended to fulfill the requirements of the GDC 24 which requires two independent reactivity control systems utilizing different design principles. The system is Seismic Category 1 and Safety Class 1 and has a scram stroke of 37.5 inches. The rods are inserted by gravity and hydraulic forces. A schematic of the drive system is shown on Figures 6 and 7 of Attacnment D. A comparison of the PCRS and SCRS mechanism is given on page 8 of Attachment D, and a summary of the diversity designed into these systems is given on pages 9 and 10 of Attachment D. Five prototype SCRS devices have been tested and a total of 3600 scrams have been performed. There have been no failures to scram and all scrams have been within the required insertion time. r i Plant Reactivity Control, R. Tinder, W/CRBR Project Mr. Tinder discussed the CRBR control system. The system is designed to q l provide automatic control in the 40% to 100% power range, and to provide i load following capability for the plant. A schematic of the control system is shown on page 11 of Attachment D. The plant control system has been subjected to human factors review and to a failure modes and effects analysis. m. 1
t. Meeting Minutes on CRBR Working Group 5 12/10/82 Design Features of the CRBR Reactor Shutdown System, C. McCrea, W/CRBR Project Mr. McCrea discussed the significant design features of the CRBR reactor shutdown systems, The shutdown system utilizes two independent and some diverse systems and is intended to maintain the plant parameters within acceptable limits for all design basis events. Designers of the systems have program results. Mr. McCrea stated that the systems comply with all applicable industry standards, the NRC General Design Criteriaand all applicable regulatory guidance. A summary of the system features are given on page 12 & 13 of Attachment D and a summary of the events which form the design basis for this system are given on pages 14 thru 17 of Attachment D. A variety of sensors are used in the reactor shutdown system. These are described on page 18 of Attachnent D. For the most part, these devices are types that are already in use i the commercial reactor industry. The exceptions are indicated on page 18 of Attachment D. It is intended that the reliability of these devices will be established by test prior to the operation of the CRBR reactor, Control and Protection Interfaces, G. Morrison, W/CRBR Project Mr. Morrison discussed the significant features of the plant protection / control systems interfaces. He indicated that the system interfaces met all applicable NRC criteria. The system uses the Westinghouse practice of shared use of control and plant safety system scasors, with buffering. Westinghouse believes there are advantages in this shared use of sensors in that additional information is available for plant control and that the number of sensor penetrations is reduced. i e
y-t Meeting Minutes on CRBR Working Group 6 12/10/82 A summary of the protection control interfaces is given on pages 19 & 20 of Attachnent D. ~ l Status of ICSB Review, G. Mach, NRR The NRC Staff began 1.he current review of the plant protection system and the instrument control system portion of the CRBR design of late 1981. They are using consultants from EG&G in this review. It is expected that the SER on this portion of the review will be completed in March 1983. I Meeting Adjourned. NOTE: Additional meeting details can be obtained froma transcript cf this meeting available in the NRC Public Document Room, 1717 H Street, N.W, Washignton, D.C., or can be purchased from Washington, D.C. 20024, (y; Inc., 400 Virginia Avenue. S.W., Alderson Reporting Compan 202) 554-2345.
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g i ns ne. ~ AsMoory Committee on Reestor Safeguards. Subcommittee en Censh esmoultants, amt Staff. 7Woons aseMag 5tiver greeder Reactor Wortirig Group to make ers! statements aboeld asefy Ibe Cognisant Federal Employee as for en Structures and Idsteriale. 80eeting as practicable so that appropriate & ACRS Subcommittee on Clinch arrangents can be made to aHow the River Breeder Reactor (CRBR) Working necessw time the mung for Group on Systems Integration and such statmak Instrumentation Control. September 30. Pchlic attendance. g win 6 open to N enum utin 1982. Room 1066.1717 H Street. NW. Washington.DC.N Subcommittee will ht forsubject mesdag sher discuss the CRBE plant protection and beas um instrument and control systems. N**
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In accordance with the procedures outlined in the Federal Register on g, Nel [eey he September 30,1981 (46 yR 47903. oral or wntien statements may be prese)nted by esasultant who me mt may anchesse pretwauy v6rwe iogenhas members of the blic recordings wSI matters to be escmulered dartas the halenes be perx'tted during those portions of the mesmas. of me meeting n a transcri t is being h subosondtise will them hear kept, and questions may be as ed only smestasons by and hold A-weih by members of the Subcouunittee.its Q 8y hk
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hterested persons regarthag this soview. FWtherinfonnation regarding topios to be discussed, whether the meeting has been cancelled er reacheduled, se Chairman's ruling on requests for the opportunity to present oral statements and the time allotted therefor can be obtained by a prepaid telephone can to the cognhent Designated Federal Employee. Dr. Richard Savio (telephone 202/634-3267) between 9:15 a.m. and 8.130 p.m e.d.L Deted September 8.1 sal pha c steys Advisory Carnalare Manngement Oficar. pm on. swems ru.d e-maa ne sul anAme caos rmws
d. g. i EE ' hl Register / vol, gy, pg g f F. Septetnber 3, m / Mme omr8 *asi REGULA10RY 0000fdtSSION AiMeory Committee on Reactor gefoguartie. Subcommittee on CIhoh fifver Breeder Reactor, WortclRO Group on Systema integration end anstrumentetson Control; Correction & ACRS Subcouranittee title has been corrected te meeting of the Clinch River Breeder Reactor (CRBR) Working Group on Systems Integration and Instrumentation Control scheduled for September 30,1982. Room 1046.1717 H Street. NW., Washington. D.C. h Subcommittee wi3 discuss the CRBR plant protection and lastrumeat and control sys' ems. 1 All other items regarding this meeting semsin the same as announced f.- the Federal Registerpublished Monday. Geptember 13.1982 (47 FR 40200). Further information regarding topics to be discussed, whether the meeting has been cancelled or rescheduled, the Chairman's ruling on requests for the j opportunity to present oral statements and the time allotted therefor can be obtained by a prepaid telephone call to the cognizant Designsted Federal Employee. Dr. Richard Savio (telephone 302/634-3287) between 8:15 a.m. and
- ak00 p.m EDT.
Deted: September 14.1ss2. Samusil. Chuk. Acting Advisory Coa'mittee Management Officer. lrt Dec 82-25 3 Faled Mf-E e4s Nl a= '== coot nas4ws r 't:
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Irdec ration AND InStrumenyafi04 Ge+re! ATTENDEES PLEASE SIGN BELOW (PLEASE PRINT) NAME BADGE N0. AFFILIATION '10lR0//4 Ri %M MSN bi ~ 1 2 feier-Cm DOE;NKMF-PC \\ 3 @k W.[l/. o Oo e //-l a 4 /P.J.7in/?/ c//Ad-w ~ 5 PIfm>nsa EssL sta zod' ozoe. \\ 6 MtM 1JL/ S litECTI JONAu($ l 7 N ScHA'DHam Mistweenooss-8 bI A) b U3 n bfM bdm 9 V m ff?'kro W ' f ! W !A'? 10.0, W. O Al l /YL $ W E E T/4) 6 //0 NS /E / 11 G. L. MoRRt5od (AJES17N6Ho0SE 12 >. L, K'//J ': /JA C
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~ 'klE_100 v.m. MEETING ROOM U N ' 1 DATE G' fA ADVISORY COMMllTEE ON REACTOR SAFEGUARDS MEETING CRBR wor <ina Grouo on Sus 4 ems u in4ec ration AND lnstrutnenfRdi6tfbd rb\\ 4 ATTENDEES PLEASE SIGN BELOW i (PLEA.SE PRINT) l NAME BADGE NO. AFFILIATION YMIod SrL AI!L16MILPD 1 T&uu eHem n n s DavcA & moran uitc/ceteen I 4 P' B, I-lo.l z_ et n 5 1A)* Lw sb 6 f. lO-7 l.). Y-om j s s/,p A : g 7 / 6--,:A 10 D, umW 11 5 ld -, 12 )Od n/ 13 I ~ 16 i 17 18 19 l 20 i i 1
wa. ' A77?kW/M&b L ~ 9/15/s2 ACRS Working Group on Systems integration and j Instrumentation and Control Washington, D.C. September 30, 1982 - PROPOSED AGENDA - "CRBR Plant Protection and Instrumentation and' Control" Presentation Actual Speaker Time Time 1. Executive Session 15 Min. 1:00 - 1:15 pm 2. NRC Staff Presentation 30 Min. 1:15 - 1:45 pm A. Status of the NRC Review B. Schedule for Coepletion of the NRC Review 3. Technical Presentations by Applicant A. Introduction C. Clare 15 Min. 1:45 - 2:00 pm B. Reactivity Control D. Dontals 30 Min. 2:00 - 2:30 pm
- BREAK ***
15 Min. 2:30 - 2:45 pm C. Reactor Control Mechanisms
- 1) Primary Control Rod G. Smith 30 Min.
2:45 - 3:15 pm System
- 2) Secondary Control Rod R. Lawrence 30 Min.
3:15 - 3:45 pm System D. P1 ant Control /P1 ant Pro-tection System
- 1) Description of Plant R. Tinder 45 Min.
3:45 - 4:30 pm Control System
- 2) Description of Plant G. McCrea 1 Hr.
4:30 - 5:30 pm Protection System
- BREAK ***
15 Min. 5:30 - 5:45 pm
- 3) Control System / Pro-G. McCrea &
30 Min. 5:45 - 6:15 pm l tection System Inter-R. Tender actions 4. NRC Staff Comments 30 Min. 6:15 - 6:45 pm 5. Subcommittee Comments and 15 Min. 6:45 - 7:00 pm Discussion of Future Schedule w
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SCRAM-RELATED FEATURES LATCHED UNLATCHED PNEUMATIC GRAVITY F RCE VE T p l PISTON / } } PISTON-- + =. 80 l R l g /( .i[fh! COLLET ~ rf HEAD -~ <A f' COLLET /s GRIPPER g CONTROL ROD )( ~~- HYDRAULIC [} FORCE y W
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7 CONTROL ROD SYSTEMS COMPARISON PRIMARY SECONDARY D g STATOR-COLLAPSIBLE SOLENOID diff ROLLER NUTS-kj. l NEUMATIC PISTON -LEADSCREW j~ ' LATCH TENSION y [h-SCRAM ROD OASPRING s DRIVELINE 1 DASHPOT _St 'hCOUPLING ~ _[ CONTROL RODdA SCRAM LATCHM fr-DAMPER CONTROL ROD-1 ({h-PIN BUNDLE PIN BUNDLE PISTON-F
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DIVERSITY BETWEEN SCRS AND PCRS DESIGN FEATURE (& LOCATION) i SCRAM OPERATIONS SCRS PCRS SENSORS AND LOGIC GENERAL COINCl-LOCAL COINCIDENCE t GENERATE SIGNALS DENCE LOGIC LOGIC TWO-OUT-OF-SCRAM VALVE SCRAM BREAKERS THREE PPS INPUTS SOLENOIDS DE-TRIP (EQUIPMENT INITIATE SCRAM ENERGlZE PANELS) (INDIVIDUAL SCRDMs) POWER REMOVAL TRIPS PNEUMATIC MAGNETIC FIELD MECHANISM CYLINDER PRESSURE COLLAPSES VENTS THROUGH (INDIVIDUAL SCRAM VALVES PCRDMs) (INDIVIDUAL SCRDMs) FORCE HOLDING TENSION ROD DROPS ROLLER NUTS CONTROL ROD IS 1/4 INCH CAUSING DISENGAGE LEAD RELEASED LATCH TO RELEASE SCREW (PCRDMs-CONTROL ROD ABOVE REACTOR COUPLING HEAD VESSEL HEAD) (TOP OF CONTROL ASSEMBLY-CORE REGION) 942-M 11
DIVERSITY BETWEEN SCRS AND PCRS (CONT.) DESIGN FEATURE (& LOCATION) PCRS SCRS SCRAM OPERATIONS SCRAM ASSIST FORCE SODIUM FLOW SCRAM SPRING EXERTS FORCE ON ACCELERATES CONTROL CAUSES NET DOWN-ROD DOWNWARD WARD FORCE ON DRIVELINE SCRAM ASSIST (ELEVATION ABOVE PISTON (BOTTOM OF REACTOR VESSEL MOVABLE CONTROL HEAD) ROD) CONTROL ROD MOVES ALL MOTION DRIVELINE lNTO ACTIVE CORE OCCURS BELOW ATTACHED TO CORE' OUTLET CONTROL ROD REGION MOVES THROUGH (UNLATCHING REQUIRED 1/4 INCH REACTOR UPPER MOTION THROUGH INTERNALS UPPER INTERNALS STRUCTURE STRUCTURE) CIRCULAR BUNDLE HEXAGONAL BUNDLE INSERTS INTO A INSERTS INTO A CIRCULAR DUCT HEXAGONAL DUCT SS2-295412
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N-f I REACT 09. SHUTDOWN-SYSTEM DIVERSITY
- i PRIMARY SYSTEM SECONDARY SYSTEM CONTROL ROD INSERTION GRAVITY WITH SPRING ASSIST GRAVITY WITH HYDRAULIC ASSIST RELEASE CIRCUIT BREAKERS IN 2/3 2/3 SOLENOID OPERATED PNEUMATIC
,'l3 ' ARRANGEMENT VALVE LOGIC LOCAL COINCIDENCE GENERAL COINCIDENCE ISOLATION l!GHT EMITTING DIODE DIRECT COUPLED ELECTRONIC CIRCUITRY INTEGRATED CIRCUITS DISCRETE COMPONENTS i MAIN CABLE TERMINATION UPPER CABLE SPREADING ROOM LOWER CABLE SPREADING ROOM INSTRUMENTATION COMPENSATED ION CHAMBERS Ft.SSION CHAMBERS PRESSURE S SPEED FLOW STEAM & FEEDWATER FLOW STEAM DRUM LEVEL r't i ~
_1...._. ~..__-- E. l._. ~.-.. i) Q glARD / / CRBRP REACTOR SHUTDOWN SYSTEMS PRIMARY SECONDARY n SENSORS T T T [' TRANSMITTERSl l I )l I l 1 i -COMPARATORS l [ ] [ 7 ( 3 ' // M I 24 11 24 11 24 l' l .ooic(1/is]I[1/is)I[1/1s] L t 3 11 9if Jli iff 11f fit l 1/24 ' [ 1/24 ] l 1/24 l, I -'.q fgl b g a s H + H + SCRAM 4 24 BREAKERS RODS l POWER TO 9 PRIMARY RODS I t U l TO HTS BREAKERS \\
o t I h TABLE 7.2-2 ~ s PPS DESIGN BASIS FAULT EVEhTS I t Prl--- Secondary Reactor 4 Fault Events Shutdown Svstam Stetdown Swstem I. An*Icloated FanIts A. ReactiyIty DistorbancesW Positive Rampt 15d/sec and Steps 510 Startup Flux-Delayed Flux or Ste tup Nuclear Flux-Pressure 5-40% Power Flux-Delayed Flux or Modified Nuclear Rate or Flux-Fressure Flux-Total Flow 40-100$ Power Flux-Pressure Flux-Total Flow Full Power High Flux Flux-Total Flow Negative' Ramps and Steps Flux-Delayed Flux ModifiedNuctORate B. Sodium Flow Disturbances 7-' Coastdown of a Single Primary or Primary-intermediate Prfr,ary-intermediate Intermediate Pump Speed Mismatch Flow Ratio Loss of 1 HTS Loop Flux-Pressure Primary-intermediate Flow Ratto I Loss of 3 HTS Loops HTS Pump Frequency Flux-Total Flow ( 5,.5 Ra I 5?! I ~ / 4 I
TABLE 7.R-2 (Continued) Fault Events primary Reacter Shutdoun 51 stem Secendary Aeactor Shutepun System i, C, Steam Side Disturbances Evgporator Module Isolation Valve IHX primary Outlet Evaporator Outlet Me -l C194pe Temperature Temperature S Stea> Feeduster Flow Evaporator Outlet Me (ygsrheaterModuleIsolationValve logpe Mismatch Temperature Estar Side Isolation and Dump IHX Primary Outlet Evapprator Outlet te of Single Evaporator Temperature Temperature Water Gide Isolation and Dump Steam-7eeduster Flos Evaporator tuttet h of Single Superheater Mismatch Temperature Water Side Isolation and Dump of Steam-Feeduster Flow Evaporator Outlet Es Soth (veperators and Superheater Mismatch Tamperature Steen Drum Level #g:. (ess of Normal Feeduster Steam-Feeduster Flow . s.g::Q.: ppg.y. .,.py Mismutch ..e Turbine Trip with Reacter Trip-Steam-Feeduster Flow Steen Drum'Le6el .d I (Loss of Main Condensee se Mismatch Similarproblem) inadvertent Opening of Eveperetop Steam.Feeduster F19e Steen Drum Level l Oullet Safety Velv4 Miesetch j Inadv.ertent BMning of Superheatee Steam.Feeduster Flow Steen Drum Level i Outlet Safety Valv4 Micast4h ) InadieTetent Opening of Eveseretor INR Primary Outlet Evaporater Outlet ne Inlet DiEI Vild .famperature Temperstere P 1 9 I 6 t5 J. i
~~~ ~ T,, - -- @ARD TABLE 7.2-2(Continued) Fault Ewee n Primary si,4ctnr Shutdene Systep secondarv Reacter Shutdeun System. II. Unlikely Faults g II) A. Ray,tivity Cinturbances Pos.itive Ramos $54/sec and Steps $404 5t, pts. Plus. Delayed Flus or Startup Nuclear Flus.oressure t 5-AGE pense Flus Belayed Flus er ' Modified Nuclear Bete er Flux-Tetal Fleu j py,,,,,,,,,,, % 1005 Pease Flus. Pressure Flum-Tetal Fleu Fyll peute Nigh Flus Flus Total Fi m 59 g Fies 9(eturtenegg di S.- Intermediate Primary.I' M Ples Primagi=ter-nati..g.g.u Eclapy Pump teleure sned Intermediate Primeppeletoft digte M Inteistdiets Pug teleure Primaryisastch Regie 5peej n III C. Steam Side Notectenoeg Steae Lint treet Stees Foodneter Fles tveperator Outlet No Miametch ?sepereture i R'cifcoittien kies treet Steam Feedseter 914e Steen Drum Level 1 e ~ m oneten Steen tr e Level Fe~edBEtW k.illt treet , Steam *Feeduster Flow i l 't
s r a E ARD TAstt 7.2-2 (contined) Fault Events; i Primary lle.)_Ctor Shutdam Erstem g d Fei. lure of Steam thsip system Steam-Feeduster Flow Steen Drum Level Mamatch S94.1 p Water teettien in SteamIII Steam Feeduster Flow Sodium-Water Reaction [ bAaret9P M gattch t II[. Extremely Unlikely I A. Re4ctivity Disturbances Positive Reses 1 it.0/see Startup Flus-Delayed Flux Stcrtup Nuclear- %= del Pome, Flus. Delayed Fles or Rdified Nuclear Rate or Flus. Pressure Flum-TotalF19jt i 46 1005 Peuer Flus.Fressure Flum-Totek1 h ~,l_A ~. - ' 'y Full Pomer Nigh Flus Flus-Total Flow i i i (1) T,hemaxianinantielpetedreactivityfavltr maximum insertion rate of apprestostely d.geults from a sinele failure of the control system with a 1 cents per tecend. j (2) TAmanjinan unlikely reactivity faulle result from smittele control system falleres' leading to with-drawal of six rods at no6imi speed se one red at the monimum mechanical speed. (3) The PPS is figuired to terihinate the results of these estessely unifkely events within the umbrelle l thiisieni sjEcified as emergency for the design of the sejer components. 'I /1
~ g. M ARD REACTOR SHUTDOWN SYSTEM SENSORS W oGM ~ LOCATION SENSOR TYPE ~ 3 e a~4. g===- =7 - p w.w = PRIMARY &.= M.Lg NUCLEAR Flux COMPENSATED ION REACTOR CAVITY WALL CHAMBER INLET PLENUM NAK TRANSMISSION v/ INLET PLENUM PIPING PRESSURE SENSOR PRESSURE PRIMARY PUMP IACH0 METER PUMP SHAFT SPEED INTERMEDIATE TACHOMETER PUMP SHAFT PUMP SPEED PUMP ELECTRICS UNDERFREQUENCY INTERMEDIATE PUMP RELAYS STEAM FLOW VENTURI WITH DP SUPERHEATER DUTLET PIPE SENSOR FEEDWATER FLOW VENTURI WITH DP STEAM DRUM INLET PIPE SENSOR REACTOR VESSEL INDUCTIVE PROEE / REACTOR VESSEL SODIUM LEVEL IHX PRIMARY CR/AL THERMOCOUPLE PRIMARY IHX OUTLET OUTLET TEMPERATURE SECONDARY NUCLEARFLUx FISSION CHAMBER REACTOR CAVITY WALL PRIMARY PUMP FL0w PERMANENT MAGNEI v' PRIMARY COLD LEG PIPE FLOWMETER f Ft r INTERMEDIATE COLD l INTERMEDIATE PUMP PERMANENT MAGNET LEG PIPE FLOWMETER FLOW STEAM DRUM STEAM DRUM LEVEL DP.' S'E,NSOR i EVAPORATOR OUTLET CR/AL THERMOCOUPLE EVAPORATOR SODIUM OUTLET j SODIUM TEMPERATURE z bNDERVOLTAGE RELAYS PRIMARY PUMP PUMP ELECTRICS REACTION PRODUCT SODIUM WATER DP SENSOR DUMP LINES REACTION W.
lAR PROTECTION / CONTROL INTERACTION l SENSOR CHANNELS MEDIAN CONTROL SYSTEM RESPONSE PROTECTION SYSTEM A(II B(21
RESPONSE
C t' POWER RANGE H L N N NORMAL NOT REQUIRED Flux L H N N NORMAL NOT REQUIRED H H N H Flux CONTROL: DECREASE IN REACTOR NOT REQUIRED i POWER TEMPERATURE CONTROL: INITIAL DECREASE NOT REQUIRED IN REACTOR POWER FOLLOWED BY PARTIAL / TOTAL RECOVERY L LN L Flux CONTROL: INCREASE IN REACTOR NOT REQUIRED POWER LIMITED BY ROD BLOCK CIRCUITS (SECONDARY FLUX) l TEMPERATURE CONTROL: INCREASE IN NOT REQUIRED REACTOR POWER LIMITED BY ROD BLOCK CIRCUITS OR TEMPERATURE FEEDBACK LOOP I (1) CHANNEL A IS ASSUMED TO BE UNDERGOING TEST. CHANNEL IS TRIPPED DURING TEST. (2) CHANNEL B IS ASSUMED TO EXPERIENCE FIRST FAILURE. SIGNAL DEVIATION IS ASSUMED TO BE INSUFFICIENT TO CAUSE A CHANNEL TRIP. 19 .e
PROTECTION / CONTROL INTERACTION (CONTINUED) I SENSOR CHANNELS MEDIAN CONTROL SYSTEM RESPONSE PROTECTION SYSTEM
RESPONSE
A(l) B(2) C i WIDE RANGE H L N N NORMAL NOT REQUIRED FLUX L H N N NORMAL NOT REQUIRED l H H N H SPURIOUS ACTUATION OF ROD BLOCK NOT REQUIRED l CIRCUITS l L L N L FAILURE OF ROD BLOCK CIRCUITS NOT REQUIRED l j PHTS ScDlum H L N N NORMAL NOT REQUIRED L H N N NORMAL NOT REQUIRED L L N L SPEED / MANUAL FLOW CONTROL: NORMAL NOT REQUIRED AUTO FLOW CONTROL: INCREASE FLOW IN NOT REQUIRED 4 ONE PRIMARY LOOP H H N H SPEED / MANUAL FLCW CONTROL: NORMAL NOT REQUIRED AUTO FLOW CONTROL: DECREASE FLOW IN PRIMARY RSS RESPONDS UPON ONE PRIMARY LOOP DEMAND 2b Jus .}}