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(ACR-700) Simulator - International Atomic Energy Agency

1. ee re The screen shows the parameters relevant to controlling the inventory in the reactor coolant loop Inventory control is achieved by controlling pressurizer level lt Reactor Trip Turbine Trip e Pressurizer level is normally under computer control with the setpoint being ramped as a function of reactor power and the expected shrink and swell resulting from the corresponding temperature changes The screen provides a PRESSURIZER LEVEL CONTROL section showing the CONTROL mode AUTO or MANUAL current PRESSURIZER LEVEL m PRESSURIZER LEVEL SETPOINT m Level control may be transferred to MANUAL using the mode control pop up and the SETPOINT can then be controlled manually using MANUAL SETPOINT pop up NOTE in order to control the pressurizer level MANUALLY one must use the pop up menu to switch the control mode from AUTO to MANUAL first then the level setpoint value will be frozen as shown in the numeric value display Observe the display message below the MANUAL SETPOINT button If it says MAN O P that means the level setpoint can now be controlled by the MAN pop up menu If it says MAN O P NOT OR that means the MANUAL setpoint signal from the MAN pop up and the frozen setpoint value do not match One must then use the MAN pop up menu to enter a value equal to the frozen numeric value display t
2. 17 54 2 20 11PM E 17 54PM 2 20 11 PM Reactor Power RIH 1 RIH 2 Pressure 110 Tj F el 2 20 11 PM 17 54 PM Resolution Time Scroll EE e Reactor Reactor Generator Output Coolant Core Main STM Press Run Iesel System w Neutron Pwr Thermal Pwr Average ak is Pressure Flow kg s BOP STM Flow cc E We E TE pp ap Hi Neut Pur Log Main SmpresHi Stm GenLevelH Loss RcPmo s PRZR RCTR OUT Pressure BLEED CDSR Level PRZR Level amp Setpoint PRZR Spray Flow Bleed Cooler 2 20 11 PM BLEED CDSR Pressure SS 2 20 11 17 54 2 20 11 Time Scroll FLOW DIAGRAM NOTE Screen only shows typical arrangements De Out In ROH PRESSURE CONTROL MODE DI Norma Not all valves are shown final design may change Inventory Reactor Reactor Generator Output Primary Coolant Core Main STM Press 2440 3 Neutron Pwr Thermal Pwr Average ROH Pressure Flow kg s BOP STM Flow Iterate amp Pressurizer w mme FWFI low RECH REG KE 118 L Sten Backread LL setback Redd L Turbine Runback _ _ Lass ees Casse EE EE Hi Neut Pwr LogR ROH Press Hi Main Stm Pres Hi Stm
3. steam turbine stop valves MSV status Condenser steam bypass dump valves status and open Generator output MW is calculated from the steam flow to the turbine e Condenser and condensate extraction pump are not simulated Simulation of the feedwater system is simplified the parameters displayed on the plant overview screen are Total feedwater flow to the steam generators kg sec gt Average feedwater temperature after the high pressure heaters HPHX gt Status of boiler SG feed pumps BFP is indicated as red if any pumps are ON or green if all the pumps are Note that while the simulator is in the Run mode all parameters are being continually computed and all the displays are available for viewing and inputting changes Note to facilitate simulator users a better understanding of the ACR 700 technical features a hot button is provided for users to navigate the five unique features of ACR 700 a Horizontal Fuel Channel b Fuel Bundle Design c D20 moderation d On Power Refueling e Passive Safety 29 3 6 control loops Neutron Thermal Turbine PWR FP PRIMARY COOLANT e Seam T Eeer Valve Normally Open 120 0 ont Steam A Discharge Valve Normally Closed 100 0 Pressure 2 Si 50 0 25 0 fem Flow Measurement Governor Valves Generator Steam Generators Pressurizer Level m Com A Rm u
4. 0 0 2 0 4 0 12 37 36 PM 12 39 09 12 37 36 PM 12 39 09 ZCU s amp MCA s Avg Pos Reactivity Change Delta mk D 0 Chambers 12 37 36 PM Resolution Top Bottom 0 079 FLUX e 59 Side T omg SE 0 00 d Front Back 0 145 Out In Reactor Power Reactor Reactor Generator Output Primary Coolant Core Main STM Press 1 Neutron Pwr Thermal Pwr Lnd Pressure Flow kg s BOP STM Flow Lee Ls Flow E 109 Transients with additional malfunction both Reactor Setback Stepback failed Turbine Trip ROH Press Lo Lo Step Back Req d Setback Req d Turbine Runback Gen Breaker Opn Coolant Flow Lo Stm Gen Level Lo Low Fud Pur Trip Main BEP S Trip _TOMAIN STEAM HEADER dea HTS process parameters at Pump Discharge 12 58 58 PM 1 01 13 12 58 58 PM 1 01 13 e ROH 1 ROH 2 Pressure Feed Bleed Flow UJ ROH 1 emm O d elass 0207 Se een gl 13171 HD 12 58 58 1 01 13 12 58 58 PM 1 01 13 RIH 1 RIH 2 Pressure Reactor Power M 14000 110 Ke 13500 75 3045 F 3356 4 50 gl 12075 12500 AVG CORE FLOW 5737 18 PM 1 01 13 COOLANT Tavg 255 74 daa 7 Resolution Time Scroll e
5. 110 5 18 Loss of 2 PHT pumps This malfunction event is a more serious accident than that described in Section 5 17 because of drastic reduction of coolant flow in losing two PHT pumps gt Go to reactor coolant system screen insert the malfunction for Loss of 2 PHT pumps in Loop 1 Observe that PHT Pumps 1 and 2 are tripped off and the coolant flow is decreasing rapidly Observe coolant flow in the other PHT pumps Observe that reactor power is stepped back Record the reactor power after the malfunction is initiated Observe the coolant pressure and temperature transients bleed and feed flow transients reactor power thermal power and turbine power transients Repeat the malfunction event again with the use of the reactor coolant system screen but before doing so first insert the malfunction for Loss of PHT Pump PI Then on the reactor coolant screen manually turn off PHT pump P3 The purpose is to study how the system thermal margin is challenged without two PHT pumps on the same loop supplying primary coolant to RIH 1 As well what reactor protection is activated to mitigate the event Observe the reactor power transient coolant pressure and temperature transients reactor neutron power reactor thermal power turbine power transients Describe and explain the difference in responses when compared with the previous malfunction transient Repeat the malfunction event again with the use
6. signal the RWS injection valves to the reactor inlet headers are open When the HTS depressurize the water from the RWT could directly be injected to the reactor inlet headers e Soon after ECC injection and steam generator crash cooldown begin emergency coolant water begins to refill the core pass As a result fuel and sheath temperatures start to decrease e The refills both core passes and a quasi steady state flow pattern is established e Long term cooling is maintained by the flow of ECCS coolant through the circuit with decay heat removal by the ECCS heat exchangers and through the break For details of the ECC Flow Diagram press the button Emergency Core Cooling Flow Diagram shown at the bottom right corner of the screen 71 4 ACR BASIC OPERATIONS amp TRANSIENT RECOVERY 4 1 Plant load maneuvering reactor lead POWER MANEUVER 10 power reduction and return to full power 1 Initialize the simulator to 100 FP 2 Select ACR reactor power control screen 3 Run the simulator by pressing the run button 4 Select the plant mode to be REACTOR LEAD 5 Record in Table II the following parameters in the full power column before power maneuvering TABLE PLANT LOAD MANEUVERING REACTOR LOAD Parameter Unit 2 3 4 Comments 90 just 90 return to reached stabilized 100 stabilized Reactor Neutron Power Reactor Thermal Power Reactor Power SP Act
7. Max Out Max In FLOW DIAGRAM Reactor Coolant Reactor Reactor Generator Output Coolant Core Main STM Press erate Syste m ow Neutron Pwr 96 Thermal Pwr Average KE Flow kg s BOP STM Flow ERU a FW Flow NET eT unes Reactor Trip Turbine Trip Step Back Req d Setback Req d Turbine Runback Gen Breaker Opn 65 Coolant Flow Lo Stm Gen Level Lo Low Fwd Pur Trip Main BFP S Trip Hi Neut Pwr LogR ROH Press Hi Main Stm Pres Hi Stm Gen Level Hi Spdr Gear in Man 1 Reactor Pwr amp Thermal Main Steam Hdr Pressure amp SP PLANT MODE REACTOR LEADING D TRML i SP P POWER RATE amp TARGET LOAD CONTROLLED VARIABLE ee gee RAKE TARGET LOAD D 68 27 100 00 5 TO 100 1 01 13 PM 12 58 58 PM 1 01 13 PM Current Target Load amp Turbine Pwr SG amp 562 Boiler Level 0 10 0 01 T01 POWER RATE S IT 0 19 STEAM GENERATOR PRESSURE SETPOINT CONTROL aa emilka sPMopE DI oo 6400 sp kea 1 01 13 PM 12 58 58 PM 1 01 13 PM Resolution Time Scroll Cm H Max Max In BECH Reactor Generator Output Coolant Core Main STM Press zm Neutron Pwr Thermal Pwr Average ROH Pressure Flow kg s BOP STM Flow erate kPa z eem S FW Flow Sy TT ET ES 1 EDD
8. 2 02 45 Governor Position MAN SP NOT OK MSSV1 Combined TURBINE TRIP STATUS MSSV2 MSSV s flow k TRIPPED E Speeder Gear e 000 SPDR GEAR CONTROL AUTO H TURBINE RUNBACK D B SE gem Ca E Resolution Time Scroll FLOW DIAGRAM Max Out Max In NOTE Screen only illustrates main steam arrangements MAN SP NOT OK asov F oa Not all valves are shown See Flow Diacram for details ACR Turbine Reactor Reactor Generator Output Primary Coolant Core Main STM Press EX G tor Y Neutron Pwr Thermal Pwr nns Flow kg s STM Flow Sg Tterate FW Flow SC RE RECH CT RENE AERE Reactor Trip Turbine Trip Fuel Temp 114 Step Back Redd Ven Hi Neutron Pwr ROH Press Hi Hi Coolant Flow Lo Stm Gen Level Lo PRZR Lvl Hi Low Fwd Pwr Trip Main BFP s Trip Hi Neut Pwr LogR ROH Press Hi Main Stm Pres Hi Main Steam Hdr Pressure amp SP PLANT MODE REACTOR LEADING D SP P POWER RATE amp TARGET LOAD CURRENT OPERATOR CONTROLLED VARIABLE TARGET INPUT TARGET RANGE TARGET LOAD D 0 00 100 00 5 TO 100 0 0 0 01 03 PM 2 03 20 PM E 1 03 2 03 20 POWER 5 010 usd Current Target Load amp Turbine Pwr SG amp SG2 Boiler Level STEAM GENERATOR PRESSURE SETPOINT CONTROL 3i53 keA
9. As well observe the feedwater flow to SG 1 Observe changes in primary coolant pressure and the surge flow from pressurizer Reactor setback will occur first on low SG 1 level Reactor trip will occur on low low SG 1 level Observe the coolant pressure transient and the surge flow from pressurizer Observe level in SG 1 gt As reactor is tripped SG pressure is dropping rapidly causing the turbine governor to runback the turbine that is closing the turbine governor control valves This results in rapid reduction of MW to zero leading to turbine generator trip on zero forward power ASDV 0 0 Neut Thrm Pwr Turb Pwr D A stm flows Extraction Main _ ES gt coesu Wwe sc Lv se v 11 6463 Main Steam Header 3 22 01PM E 19 43 PM 3 22 01PM 5 Ext Stm Flows SG Level m 561 562 Time Scroll Resolution 4 A Max Out In NOTE Screen only illustrates main feedvarer amp extraction steam flow arrangements Not all valves are shown See Flow Diagram for details Feedwater amp Primary Coolant Main STM Press bd Average ROH Pressure BOP STM Flow Iterate a e P fon EE RER CT TO acm Generator Output Reactor Reactor Core Neutron Pwr Thermal Pwr Flow kg s 88 5 3 FW LCV 1 fails
10. 3 59 07 PM 56 44PM 3 59 07 PM RIH 1 RIH 2 Pressure Reactor Power RIH1 14000 110 35758 Mi Resolution Time Scroll e Max Out Max In Main STM Press Run eae BOP STM Flow c ES LL E TT s MM Mal Reactor C Reactor System Neutron Pwr Thermal Pwr Average ROH Pressure Flow kg s Reactor Generator Output Primary Coolant Core RCTR Neut Thrm Pwr Generator Output MW Main Steam Header P MSV Inlet To HP heaters OUTPUT gross 0 00 Mw sg 56 44 PM s n as Turbine Speed 3 59 07 PM Steam CSDV ASDV MSSV Flow ep HE v Ul Em 600 0 400 0 200 0 EGEMAPM z 3 59 07 PM MSV Inlet Pressure 56 44PM 3 59 07 PM To Governor Position ernor Position Feedwater 105 07 System 80 0 EN Combined TURBINE TRIP STATUS 60 0 A EC weu Seen se eeder Gear n 9 0 00 k 20 0 j Geet AUTO E 56 44 PM 3 59 07 PM Son z erm TURBINE Resoluti Time Scroll RUNBACK D Resoluton 4 3 FLOW DIAGRAM Max Out In NOTE Screen only illustrates main steam arrangements Not all valves are shown See Flow Diaaram for details Reactor Generator Output Primary Coolant Core Main STM Press Eet Re Neutron Pwr Thermal Pwr Average ROH Pressure Flow
11. MODERATOR TEMPERATURE reactivity feedback 0 024 MK deg C from 70 to 90 deg C COOLANT TEMPERATURE reactivity feedback 0 01 MK deg C from 290 to 310 deg C GADOLINIUM reactivity feedback 1 ppm will yield 6 MK Addition rate 0 5 MK minute removal rate 0 1 MK minute FRESH FUELreactivity 44 MK Note that reactivity is a computed parameter and not a measured parameter at the actual plant It is displayed on the simulator as a means of understanding how the reactor is being controlled using reactivity as the parameter Note also that when the reactor is critical the Total Reactivity must be zero This screen also shows the movement of the zone control units ZCU and mechanical absorbers rods MCA as a function of the Reactor Power Error see definition below The relationship is depicted by the movement of a yellow cursor shown on a graphical X Y plot The Plot has Y axis as Average Zone Control Rods Position and X axis as Reactor Power Error and is known as Reactivity Limit Control Diagram 40 100 0 95 0 90 0 e 85 0 80 0 75 0 70 0 65 0 60 0 55 0 50 0 45 0 40 0 35 0 p 30 0 25 0 20 0 15 0 10 0 5 0 0 07 l 1 1 1 1 1 I 7 0 60 50 40 30 20 10 00 10 20 30 40 50 60 2 N E N n R 2 Reactivity Limit Control Diagram As mentioned above the mechanical absorber rods are divided into two banks The drive logic fo
12. shutdown rods known as SDSI into the core by gravity As well the liquid poison shutdown system SDS2 is also activated Note SDS2 is not simulated gt well all the ZCU and MCA rods are inserted into the core at maximum speed end result is to put lots of negative reactivity into the core such that the nuclear fission chain reaction in the core is stopped immediately This exercise demonstrates the manual reactor trip transient and how to recover and return the reactor to full power e Initialize the simulator to 100 FP e Go to ACR Control rods amp SD rods screen note the shutdown rods position ZCU rods position MCA rods position e Go to reactor power control screen record the reactivity mk contribution from the reactivity devices and the feedback effects Le SD rods ZCU rods MCA rods Xenon fuel temperature moderator temperature amp Gadolinium Manually trip the reactor using the pop up control at the left bottom of the screen e Observe the response of the overall unit Go to trends screen observe the trends for reactor power reactor coolant pressure SG pressure steam flow feedwater flow and generator power e Wait until generator power is zero and reactor neutron power is less than 0 1 e Go to ACR Control rods and SD rods screen reset the reactor trip and shutdown system SDS Observe that the SD rods are withdrawing e Record the time using the di
13. tables Comparison of CANDU 6 and ACR 700 Unit Data DATA CANDU 6 ACR 700 Reactor Type Thermal Output to Steam Generators MWth Coolant Moderator Calandria diameter m Fuel channel Number of fuel channels Lattice pitch mm Reflector thickness mm Enrichment level Fuel bum up MWd te U Fuel bundle assembly Length of bundle mm Outside diameter maximum mm Bundle weight kg Bundles per fuel channel Heavy Water Moderator Systems Mg D O Heat Transport Systems Mg D O Reserve Mg D O PTR 2064 Pressurized Heavy Water Horizontal Zr 2 5wt Nb alloy with modified 403 SS end fittings Sintered pellets of Natural UO 0 71 w ZU 7 500 37 element 4953 102 7 24 1 includes 19 2 kg U PTR 2034 Pressurized Light water Heavy water 52 Horizontal Zr 2 5wt Nb alloy with modified 403 SS end fittings Sintered pellets of slightly enriched UO amp Natural UO in central element Average 2 1 wt U 235 in 42 elements central element NU with Dysprosium 20 500 43 element CANFLEX 22 7 includes 18 kg U 12 DATA CANDU 6 ACR 700 Heat Transport System Reactor outlet header pressure MPa 9 Reactor outlet header temperature C Reactor inlet header pressure MPa 2 Reactor inlet header temperature Reactor core coolant flow total Mg s Single channel flow maximum kg s Steam Generators Number Type Steam temperature nominal Steam qual
14. 4 32 58 PM RIH 1 RIH 2 Pressure Reactor Power RIH amp 1 14000 RIH 2 13500 Resolution rege d Max Out Max In FLOW DIAGRAM Reactor t Reactor Reactor Generator Output Primary Coolant Core Main STM Press Run System lt Neutron Pwr Thermal Pwr Average on Pressure Flow kg s BOP STM Flow ECH FW Fh pL ST TT E 95 5 9 Turbine spurious trip This malfunction event is similar to the operational transient of turbine trip See description in Section 4 4 5 10 PRZR heaters 2 to 6 turned ON by malfunction This malfunction event causes reactor coolant pressure to increase due to the fact that all the pressurizer on off heaters 2 to 6 are turned on The rise in coolant pressure is offset by the pressurizer spray that will come into action once the coolant pressure exceeds a predetermined setpoint for spraying When this malfunction transient occurs Go to reactor coolant system screen observe that the pressurizer heater 2 to 6 are turned by malfunction Observe that the reactor coolant pressure increases and then the pressurizer spray comes in to cool the pressure down What is the net effect on reactor coolant pressure What happens to coolant temperature increase or decrease Explain the response Explain why the pressurizer level goes down and
15. Porcio Aesch BE BE NEZ MW Harry Wie Maly aun mmm Sein ue eda NE Ginn ois Tram Im EI OS valve p B 2 E m Low 2reasine Turoimes E H Main Du Transomer g c Morso Avg tuel emt 7 482 Generator Output s4 Frimary Coolen Main STM Fress Neutron Pat G9 Thermal Partea See BOP STM Flow BEER seem Shows a line diagram of the main plant systems and parameters No inputs are associated with this display The systems and parameters displayed are as follows starting at the bottom left hand corner REACTOR is a 3 D spatial kinetic model with six groups of delayed neutrons The decay heat model uses a three group approximation The reactivity calculations include reactivity feedback effects reactivity control and safety devices shutoff rods SOR zone control units ZCU absorber rods MCA Xenon Iodine fuel temperature moderator temperature coolant temperature Gadolinium fresh fuel reactivity The parameters displayed are gt Neutron power full power gt Reactor thermal power full power Reactor coolant main loop with four heat transport system HTS pumps two steam generators two Reactor Inlet Headers RIH 1 RIH 2 two Reactor Outlet Headers ROH 1 RO
16. SGPC ACR Passive core cooling ACR Trends 26 3 4 Simulator display common features The ACR simulator is made up of 14 interactive display screens or pages All of these screens have the same information at the top and bottom of the displays as follows e Top of the screen contains 21 plant alarms annunciations these indicate important status changes in plant parameters that require operator actions Top right hand corner shows the simulator status gt window under labview this is the proprietary software that generates the screen displays has a counter that is incrementing when labview is running if labview is frozen i e the displays cannot be changed the counter will not be incrementing gt the window displaying CASSIM this is the proprietary software that computes the simulation responses will be green and the counter under it will not be incrementing when the simulator is frozen i e the model programs are not executing and will turn red and the counter will increment when the simulator is running e To stop freeze Labview click once on the STOP sign at the top left hand corner to restart Labview click on the gt symbol at the top left hand corner e To start the simulation click on Run at the bottom right hand corner to Stop the simulation click on Freeze at the bottom right hand corner e The bottom of the screen shows the values of the following m
17. Turbine runba Hi Neutron Pwr ROH Press Hi Hi Stm Gen Level Lo PRZR Lvl Hi Low Fwd Pwr Trip Main BFP s Trip Hi Neut Pwr LogR ROH Press Hi Stm Gen Level Hi Loss RC Pmp s Main Steam Header P 3 Generator Output MW MSV Inlet H f A 4 32 58 PM Steam CSDV ASDV MSSV Flow 4 32 58 PM To Governor Position Feedwater 105 0 System MSSV1 80 0 Combined TURBINE TRIP STATUS 60 0 MSSV s flo Ces TRIPPED E Speeder Gear moron AUTO al MSSV2 30 41 4 32 58 PM Sr 4 32 58 PM Resolution Time Scroll Out In NOTE Screen only illustrates main steam arrangements Not all valves are shown See Flow Diaaram for details ACR Turbine Reactor Reactor Generator Output Primary Coolant Core Main STM Press G t v Neutron Pwr Thermal Pwr Average Pressure Flow kg s BOP STM Flow a FW Flow ol EE TT TT ee nemi Step Back Redd Setback Rec EESTI er Jee ne el mM rosse __Hineutenrtoor JL nose zeien LL spor Gearinman LL eseeg 1 22 Temp m 2 3 4 Flows to RIH s 219 4 E HTS process parameters at Pump Discharge 4 32 58 PM E EH 4 32 58 PM ROH 1 ROH 2 Pressure m Feed Bleed Flow fd 30 41 BM 4 32 58 0 30 41
18. 1 38 53PM 1 38 40 PM 1 38 53 PM 100 00 Top Bottom 0 077 Resolution Time Scroll nm Side Side 0012 Ra um 9 00 5 Front Back os 0 138 React Rea Generator Output Primary Coolant Main STM Press Reactor a Neutron Pe 99 Thermal Set ROH Pressure Flow als BOP STM Flow 4166 3 Freeze Iterate Kee FW Fon Fa Em s re AM This screen permits control of reactor power setpoint and its rate of change while under Reactor Regulating System RRS control i e in REACTOR LEADING mode Several of the parameters key to RRS operation are displayed on this page e The status of reactor control is indicated by the four blocks marked MODE SETBACK STEPBACK AND TRIP They are normally blue but will turn red when in the abnormal state gt will indicate whether the reactor is under TURBINE LEADING to REACTOR LEADING control this status can also be changed here gt SETBACK status is indicated by YES or NO setback is initiated automatically under the prescribed conditions by RRS but at times the operator needs to initiate a manual setback which is done from this page on the simulator the target value 96 and rate sec need to be input gt STEPBACK status is indicated by YES or NO stepback is initiated automatically under the prescribed conditions by RRS but at times the operator needs to initiate a man
19. 96 and neutron log rate gt 0 sec Gd will be added automatically resulting in a negative reactivity rate of 0 5 MK per minute with a delay of 30 seconds Gd in core will be slowly burnt out at a time constant of 9 hours at nominal core conditions However if needed Gd can be removed MANUALLY resulting in a positive reactivity rate of 0 1 MK minute The screen also shows the reactor core normalized flux intensity map in color e The flux intensity scale is from 0 grey color 1 2 red color e The core flux mapping is represented in a simplified manner by 18 cells with each cell representing a section of the core coinciding with the location of one zone control unit ZCU Each cell s flux intensity is represented by a color map e Axially each cell is also aligned with a section of a lumped reactor channel fuel and coolant being modeled in the simulator For the simulator there are 6 lumped reactor channels modeled Lumped Channel 1 represented by cell Z1U Z2U Z3U Lumped Channel 2 represented by cell Z4U Z5U Z6U Lumped Channel 3 represented by cell Z7U Z8U Z9U Lumped Channel 4 represented by cell ZIL Z2L Z3L Lumped Channel 5 represented by cell Z4L Z5L Z6L Lumped Channel 6 represented by cell Z7L Z8L 291 42 The coolant flows in adjacent channels are in opposite directions namely coolant at channel 1 3 5 flows in one direction to one Reactor Outlet Header coolant at channel 2 4 6 flows in
20. EE TT TT Ee r This screen permits control of station load setpoint and its rate of change while under TURBINE LEADING control mode Control of the main steam header pressure is also through this screen but this is not usually changed under normal operating conditions ACR OVERALL UNIT CONTROL MODE can be changed from REACTOR LEADING to TURBINE LEADING TARGET LOAD on selection station load and rate of change sec can be specified change becomes effective when accept is selected The OPERATOR INPUT TARGET is the desired setpoint inserted by the operator the CURRENT TARGET will be changed at a TARGET and POWER RATE specified by the operator gt Note that the RANGE is only an advisory comment numbers outside the indicated range of values may be input on the Simulator STEAM GENERATOR PRESSURE SETPOINT CONTROL alters the setpoint of the steam generator pressure controller which is rarely done during power operation Caution must be exercised when using this feature on the simulator However this feature can be used for educational study of ACR plant responses under different secondary pressure conditions change SG pressure setpoint first use the SP Mode pop up to change the SP mode from HOLD to INCREASE or DECREASE depending on new 66 pressure setpoint target After that use the pressure SP change rate pop up to enter n
21. FP Low steam generator level not implemented as Stepback parameter but implemented as Setback parameter see below Manual stepback initiated by operator target set by operator The causes for REACTOR SETBACK are Hi zonal flux gt 120 96 setback at 0 1 s to end point 60 96 Hi flux tilt gt 20 96 setback at 0 1 s to end point 20 Main steam header pressure Hi setback if gt 6800 KPa at 0 5 s to end point 10 Low deaerator level setback if 2 075 m at 0 8 s to end point 2 High moderator temperature setback moderator temp is not modeled Low moderator pump delta P setback moderator pump delta P is not modeled Hi pressurizer level setback if gt 12 M at 0 1 s to end point 2 Low steam generator level setback if lt 10 11M at FP at 0 8 s to end point 2 Note the low steam generator level setback setpoint is a function of reactor power SP 8 8 41 31 Reactor Power normalized Hi end shield inlet temperature setback end shield inlet temp is not modeled Hi bleed condenser pressure setback not implemented yet TBD Turbine trip or loss of line setback at 0 5 s to 75 Manual setback initiated by operator target set by operator Note The Stepback and Setback Parameters indicated here could be different than those in current ACR 700 design as these parameters are subject to changes as a result of latest safety
22. GENERATOR 500 40 0 20 0 9 53 4 11 56 PM Turb Steam CSDV ASDV MSSV Flow 1200 0 98 4 11 56 PM 09 53 4 11 56 PM Governor Position MSV Inlet Pressure MSSV1 Combined TURBINE TRIP STATUS MSSV2 wssys fi 4 Ceci TRIPPED Speeder Gear 80 00 SPDR GEAR CONTROL aur H 4 11 56 PM TURBINE D tion Time Scroll RUNBACK D d d FLOW DIAGRAM Max Out Max In NOTE Screen only illustrates main steam arrangements Not valves are shown See Flow Disaram for details Reactor Generator Output Primary Coolant Core Main STM Press iterate Neutron Pwr Thermal Pwr Average ROH Pressure Flow kg s BOP STM Flo KPa d nuno BEER Reactor Trio Turbine 7 0 0 059 TT 02 7 7o 92 HTS process parameters at Pump Discharge 2825 5 asa e 13125 Gr a gr Resolution 4 I Max Out e In Reactor Reactor Generator Output Primary Coolant Core Main STM Press 4054 5 Neutron Pwr Thermal Pwr Average ROH Pressure Flow kg s BOP STM Flow 7 r FWFlow Iterate 93 5 8 Turbine bypass valve CSDV fails closed This malfunction will cause the NPP to lose its steam bypass CSDV capability in the event of turbine trip On turbine trip reactor power will be setbacked back to
23. OUT Pressure BLEED CDSR Level 12150 Retr P 10 35 04 BLEED CDSR Pressure Resolution Time Scroll FLOW DIAGRAM ROH PRESSURE CONTROL MODE D NORMAL P2 O NOTE Screen only shows typical arrangements Max Out Max In Not all valves are shown final design may change Reactor Reactor Generator Output Primary Coolant Core Main STM Press 7 5 Neutron Pwr Thermal Pwr Average ROH Pressure Flow kg s BOP STM Flow amp Pressurizerw NENNEN Ps Lene Leer Less PE MODE Hi SETBACK STEPBACK TRIP D B Kd 34 20 PM 10 35 04 10 34 20 PM 10 35 04 ZCU s amp MCA s Avg Pos 96 Reactivity Change Delta K mk 10 0 10 0 Li A 0 0 15 0 GADOLINIUM CTRL Fisidri x INT Chambers 10 34 20 PM 10 35 04 10 34 20 PM 10 35 04 MODE auto Di 100 00 FP Top Bottom 0 078 Resolution Time LUX Se Side Cal vi ISO 1 Front Back 0 146 Out Max In Reactor Power Reactor Reactor Primary Coolant Core Main STM Press 1 ue Neutron Pwr 96 Thermal Pwr Average dod Pressure Flow kg s BOP STM Flow ET FW Flow M LIRE s TT gt 104 5 15 All MCA rods stuck to MANUAL This malfunction event impairs the capability of the react
24. Outlet E Purp Astor Assembly Syston inlet BEE Feedwater BEE Heer Water Mederator EE Steam fue Buda E Ganerator pu AA P Flin Mochion INNEN Tetine A Pax W MEC ESAE ES de x E Power P Turti Aw Pressure Iurbines to Grid L OTLEX NI A Generator OWORO InN i J d Channels Atomic Energy of Canada Limited AECL has developed the ACR 700 Advanced CANDU Reactor 700 as the next generation CANDU with goals of reduced capital cost shorter construction schedule high capacity factor low operating cost increased operating life simple component replacement and enhanced safety features The ACR design is based on the use of modular horizontal fuel channels surrounded by a heavy water moderator the same feature as in all CANDU reactors The major differences in ACR as compared with CANDU are e the use of slightly enriched uranium fuel 2 1 wt U 235 in 42 pins of the fuel bundle and e light water as opposed to heavy water D20 as the coolant which circulates in the fuel channels ACR 700 Advanced CANDU Reactor is a trademark of Atomic Energy Canada Limited AECL CANDU CANada Deuterium Uranium is a registered trademark of Atomic Energy of Canada Limited AECL This results in a more compact reactor design Calandria inside diameter 31 6 less than that for CANDU 6 and a red
25. SG VAULT SG VAULT nc LAC COOLING COOLING COILS COILS SG VAULT EMCLOSURE d 50 VAULT i SG VAULT LAC LAC COOLING COOLING FANS FANS Containment Cooling System 22 3 700 MW E ADVANCED CANDU REACTOR NPP SIMULATOR The purpose of the 700 MW e advanced CANDU reactor ACR 700 NPP simulator is educational to provide a training tool for university professors and engineers involved in teaching topics in nuclear energy As well nuclear engineers scientists and trainers in the nuclear industry may find this simulator useful in broadening their understanding of ACR transients and power plant dynamics The simulator can be executed on a personal computer PC to operate essentially in real time and to have a dynamic response with sufficient fidelity to provide ACR plant responses during normal operations and accident situations It also has a user machine interface that mimics the actual control panel instrumentation including the plant display system and more importantly allows user s interactions with the simulator during the operation of the simulated ACR plant The minimum hardware configuration for the simulator consists of a Pentium PC or equivalent minimum 1 7 GHz CPU speed minimum of 512 Mbytes RAM at least 30 Gbytes hard drive 32 MB display adaptor RAM hi resolution video card capable of 1024 x 768 resolution 15 inch or larger high resolution SVGA colour monitor keyboard and mouse The operat
26. adjusted by varying the lengths of the absorbers inserted into the core based on a signal from the station computer The zone controller system is designed so that during normal operation the average zone control absorber element remains in the range 20 to 80 of full insertion The zone control system is designed to perform two main functions a Bulk control i e control of power output The zone control system will provide short term fine control of reactivity to maintain reactor power at demanded level during normal operation The bulk flux control is mainly carried out by the 10 zone control rods located near the center of the reactor vessel namely Z2U Z2L Z4U ZAL Z5U 751 Z6U Z6L Z8U Z8L b Spatial control i e control of flux and power shapes The zone control system will maintain the desired global flux and power distributions by counteracting any power distortion or oscillation brought on by a space dependent reactivity perturbation In practice the perturbations can be caused by 1 fuel burnup and refuelling of channels 2 power level changes 3 changes in the heat transport system conditions 4 xenon oscillations 5 movement of absorber elements and 7 small variations in moderator poison concentration The spatial control is mainly carried out by the 8 zone control units ZCU located near the four corners of the reactor vessel namely ZIU ZIL Z3U Z3L Z7U 471 Z9U 291 b Control Abs
27. driving OUT in GREEN region will reverse direction driving IN in LIGHT BROWN region or vice versa Hence this region serves as a deadband for which the absorber rods may not move until clear demarcation in entering the GREEN region or LIGHT BROWN region is established by the relationship of Power Error 96 versus zone control rods position NOTE the 18 zone control units ZCUs are normally controlled by the Reactor Regulating System RRS in auto mode The control of ZCU can be switched to manual mode where each ZCU can be controlled individually with the control button for IN STOP OUT Likewise the two banks of absorber rods are normally controlled by RRS in auto mode The control of individual bank of absorber rods can be switched to manual mode where each bank can be controlled individually with the control button for IN STOP OUT The screen also displayed the following parameters related to the reactivity control devices Average 8 ZCUs position 96 responsible for flux tilt control Average 10 ZCUS position 96 responsible for bulk flux control Average of all 18 ZCU speeds in per sec Absorber rods MCA bank 1 speed in 46 per sec Absorber rods MCA bank 2 speed in 46 per sec en pe bor As well included on this screen is the Gadolinium Gd control system which can be used for relatively short term core reactivity control If the control system is in AUTO mode and the Power error gt 5
28. for 90 power stabilized y Explain the responses for Primary coolant pressure Coolant feed and bleed flow changes Coolant temperature Steam generators pressure Feedwater flow and steam flow ZCU and MCA rods movement Gadolinium load changes Flux tilt Return reactor power to 100 at 0 3 by using the reactor power setpoint pop up When reactor power has returned to 100 and the parameters have stabilized unfreeze record parameter values in the column 4 of Table II return to 100 stabilized Note any major difference in parameter values between column 4 and column 1 Can you explain why the differences in parameter values if any Plant load maneuvering turbine lead POWER MANEUVER 10 power reduction and return to full power Initialize simulator to 100 full power Verify that all parameters are consistent with full power operation Select the MW demand SP amp SGPC page Change the scale on the reactor PWR amp thermal PWR and current target load amp turbine PWR graphs to be between 80 and 110 percent the main steam Hdr pressure amp SP to 5000 and 7000 KPa SG level to 10 and 15 meters and set resolution to max out Record down the following parameters in the full power column 1 of Table before power maneuvering 74 Go to reactor power control screen and record down the followin
29. kg s Generator Y BOP STM Flo Ei a FW Flow EEN TT MTC 1 Be 91 5 7 atmospheric MSSVs fail open This malfunction will cause immediate depressurization of the steam generators Responding to rapid dropping of main steam pressure the turbine will be unloaded rapidly followed by turbine trip on zero forward power Reactor power will be setbacked to 75 FP upon turbine trip On the primary side the rapid drop in steam generator pressure causes the coolant temperature and pressure transients When this malfunction transient occurs to turbine generator screen observe the main steam safety relief valves MSSV opening Observe the turbine governor valve position and that the turbine is unloaded rapidly As the turbine is unloaded observe the transient of main steam pressure Does the turbine Bypass valve open in this transient CSDV first or ASDV first Repeat this transient but this time go to reactor coolant system screen first before inserting the malfunction Observe reactor coolant temperature and pressure transient Explain why ROH RIH pressures and temperatures are decreasing json nie StepBackReqd SetbackRedd turbine runba Stm Gen Levello PRIR Lvi Hi Low Fwd Pur Trip SERE RCTR Pwr Main R NELLE am Header To HP heaters STATION Dg e 4098 0 services S90 ww T 600 0 MSV Inlet
30. of the reactor coolant system screen but this time before doing so first insert the malfunction for reactor setback amp stepback both failed The purpose is to study how the system thermal margin is challenged without the initial reactor power stepback Observe the reactor power transient coolant pressure and temperature transients Describe and explain the difference in responses when compared with the previous malfunction transient Discuss the thermal margin challenge in these cases and how the safety and control systems can cope with these challenges 111 Transients for the loss of PHT P1 and P2 ROH Press Lo Lo Step Back Reqd SetbackRedd Turbine Runback Ger Breaker Opn Hi Neutron Pwr ROH Press Hi Hi Stm Gen Level Lo PRZR Lvi Hi Low Fwd Pwr Trip Hi Neut Pwr LogR ROH Press Hi Main Stm Pres Hi Stm Gen Level Hi RIH 1 2 ROH 1 2 Temp BL 2 3 4 Flows to RIH s 400 HTS process parameters at Pump Discharge 28 51PM 1 31 08 PM 1 28 51 PM 1 31 08 PM ROH 1 ROH 2 Pressure Feed Bleed Flow zs ROH 1 ROHS2 Pressure Eoded Eon fa Toa 320 65 1 31 08 PM Reactor Power RIH 1 RIH 2 Pressure 14000 13500 13000 12500 12000 Y Resolution Time Scroll AVG FUELT 170 58 CA i 62 Max Out e In FLOW DIAGRAM Reactor Reactor Generator Output Primary Coolant Core Main STM Pre
31. opposite direction to the other Reactor Outlet Header In conjunction with the flux map of the core the flow path of the reactor coolant through the core is also shown below the flux map Cold reactor coolant from the U tubes steam generators outlets enters the reactor at the respective Reactor Inlet Headers entry points RIH 1 RIH 2 The reactor coolant from the inlet headers then travels through the respective reactor core coolant channels The reactor coolant carries the heat energy from the fuel pellets as it travels through core channels and mixes with other coolant streams before leaving the reactor at the two hot Reactor Outlet Headers ROH 1 ROH 2 The parameters displayed 1 RIH 1 2 coolant inlet flow rates in Kg sec ROH 1 2 coolant outlet flow rates in Kg sec Average fuel temperature deg C Average coolant temperature at RIHs Average coolant temperature at ROHs St dec 43 3 8 ACR reactor power control Retr Pur th Pwr Tur Pwr Coolant Delta T ROH T RIH T Deg C 110 0 100 0 MODE SETBACK TRIP les 4 yo 80 0 E 60 0 MW DEMAND SETPOINT 000 en D 110 00 wl 0 00 1 38 53PM 1 38 40 1 38 53 SETPOINT ZCU s amp MCA s Avg Pos Reactivity Change Delta K mk p 100 0 10 0 KE pe REACTOR E E samom H 2 CE 2 Fissi 15 0 Chambers 38 40 PM
32. product and is largely based on the 900 MWe CANDU system e The IAEA advanced PHWR simulator by CTI from Canada which represents the ACR 700 system e The IAEA advanced BWR which largely represents the GE ESBWR passive BWR design and was also created by CTI This activity was initiated under the leadership of Mr R B Lyon Subsequently Mr J C Cleveland Ms S Bilbao y Le n and later Mr M J Harper and Mr S D Jo from the Division of Nuclear Power became the IAEA responsible officers More information about the IAEA simulators and the associated training is available at http www iaea org NuclearPower Technology Training Simulators CONTENTS 1 IN TRODUGTION EE 1 PURPOSE eegenen EE Ee bibat E fete dis 1 1 2 BACKGROUND AND HIGHLIGHTS OF DIFFERENCES CANDU VS ACR seen eene 2 2 BRIEF ACR 700 SYSTEMS OVERV IEWN eee esee ee eee setas eee eese sese eese toas eese s 8 2 T REACTOR cereos ue Ett eme ee de nee E EEN 8 2 2 REACTIVITY CONTROL UNITS dee cdo ENEE ENNERT Eres EPI PERDRE NUS 10 2 3 HEAT TRANSPORT SYSTEM ront IRAN Rem 11 2A MODERATOR SYSTEM ive tree Ure e tete ten UO c to Et eu EURO dette ria eeu ete 12 2 5 STEAM AND EEBDWATER S Y STEM e eerte taste vtr idee esee eani a ee e 14 2 6 BALANCE OB PLANT 22 t c e eti bi drm Rar ien 16 DU SABEELYSSYSTEEMS host
33. results in the following events Emergency Coolant Injection ECI System is initiated by the signal The one way rupture discs burst open at a pressure differential of 0 52 MPa The ECC piping downstream of one way rupture discs is pressurized to the heat transport system pressure Thus the ECI injection flow will begin when the pressure in the heat transport system is about 0 52 MPa less than the ECI injection pressure from the ECC accumulators ECI injection continues until the associated ECI accumulator is nearly empty Valves on ECI interconnect line between the reactor outlet headers open up on the ECI signal to assist in establishing a cooling flow path Steam generator crash cooldown is initiated 30 seconds after the ECI signal through the automatic opening of the main steam safety valves MSSV s This assists in ECC injection by further depressurizing the HTS On the ECI signal water is automatically introduced into the containment sumps from the Reserve Water Tank RWT and the LTC pumps start automatically The long term cooling LTC pumps start automatically on a high reactor building sump level signal When the ECI accumulators are nearly empty the ECI accumulator isolation valves close and the LTC stage begins by pumping water from the reactor building sump LTC 70 delivers flow to the reactor inlet headers thereby utilizing the cooling flow path already established by the ECI system On the
34. sen DI 6 0 10 0 2 01 03 PM 2 03 20 PM Resolution Time Scroll CO Max Out sa Max In Reactor Reactor Generator Output Primary Coolant Core Main STM Press Iterat Neutron Pwr 96 Thermal Pwr Average ROH Pressure Flow kg s BOP STM Flow 863 4 eu pen E ET TT TS E Reactor Trip ROH Press Lo Lo Step Back Req d Setback Req d Turbine Runback eran Main BFP s Trip Reactor Trip RIH 1 2 ROH 1 2 Temp P1 2 3 4 Flows to RIH s e 417 0 HTS process parameters at Pump Discharge 00 34 2 02 51 00 34 2 02 51 e daas Fl ROH 1 ROH 2 Pressure Feed Bleed Flow e Eau E E oum e ULM ROH 2 PJ 13050 9 00 33 2 02 51 2 00 34PM 2 02 51 PM Reactor Power RIH 1 RIH 2 Pressure s 13500 RIH 2 72559 E 3550 1 ES 12800 Mey 000 AYGCOREFLOW 717586 Saz zeen Ein zeen E 77 ERIS 23400 Resolution Time Scroll AVG FUELT 57 55 4 l 4 Las e In Reactor Coolant Reactor Reactor Generator Output Primary Coolant Core Main STM Press 1 EE System ze Neutron Pwr Thermal Pwr Average e Pressure Flow kg s BOP STM F
35. support of reactor safety Reactor setback is the ramping of reactor power at fixed rate to the setback target in response to certain process parameters exceeding alarm limits as a measure in support of reactor safety 50 The TRIP PARAMETERS for REACTOR TRIP are Trip Parameter Design Setpoint ROP FP 123 High Log Rate sec 10 HTS high pressure 12 75 MPa g or 12 45 MPa g with 3s delay HTS low pressure 10 9 MPa g 95 FP 8 35 MPa g at 0 FP linear in between Pressurizer low level m 6 5 2 95 FP 0 lt 25 FP linear in between HTS low coolant flow kg s 85 nominal flow instrumented channel Steam generator feedline low pressure 5 4 Mpa g Steam generator low level m 9 9 m subject to change Containment high pressure TBD Note The trip parameters indicated here could be different than those in current ACR 700 design as these parameters are subject to changes as a result of latest safety and design review by AECL 51 The causes for REACTOR STEPBACK are gt Two reactor coolant pumps trip target 2 FP One reactor coolant pump trip target 2 FP Heat transport pressure high initiated at P gt 12 55 MPa target 2 FP Hi zone flux initiated if zone flux is gt 115 of nominal zone flux at full power target 2 96 High log rate initiated when d InP dt gt 7 s target 2
36. the numeric value display Observe the display message above the valve control If it says MAN O P OK that means the control valve can now be controlled by the MAN pop up menu If it says MAN O P NOT OK that means the MANUAL control signal from the MAN pop up and the frozen control signal to the control valve does not match One must then use the MAN pop up menu to enter a value equal to the frozen numeric value display then the message will say MAN O P OK The last section on this screen makes provisions for changing the reactor outlet pressure setpoint and the pressure control mode for the heat transport system The current reactor outlet pressure is shown and the reactor outlet pressure setpoint KPa may be controlled manually via the control pop up provided As well a ROH PRESSURE CONTROL MODE control pop up is provided to facilitate the heat transport coolant pressure to be controlled in two modes NORMAL or SOLID SOLID mode represents the condition that the pressurizer is isolated from the heat transport circuit meaning that the isolating valve MV1 will be fully closed Therefore in SOLID mode there will be much pronounced pressure effects increase or decrease with changes in coolant mass inventory This mode is usually used during plant shutdown or cold startup when a fast coolant pressure decrease or increase is required In NORMAL mode as usually the case in normal plant operation the isolating valve MV1 is
37. the reactor developer for CANDU and ACR However a brief system overview is presented and details are provided where necessary to describe the functionality and the interactive features of the individual simulator screens which relate to the specific ACR subsystems The user manual covers basic NPP plant operations like plant load maneuvering and trips and recovery e g turbine trip and reactor trip In addition it covers plant responses to malfunction events Some malfunction events lead to reactor trip or turbine trip Other serious malfunctions e g LOCA lead to accident situations causing actuation of the passive core cooling safety system It should be mentioned that the equipment and processes modeled in the simulator represent realistic ACR characteristics However for the purpose of the educational simulator there are necessary simplifications and assumptions made in the models which may not reflect any specific reactor vendor s design or performance Most importantly the responses manifested by the simulator under accident situations should not be used for safety analysis purposes despite the fact that they are realistic for the purpose of educational training As such it is appropriate to consider that those simulator model responses perhaps only provide first order estimates of the plant transients under accident scenarios 1 23 Background and Highlights of Differences CANDU vs ACR Host Tronsgort System
38. this screen respectively for the Bank 1 and Bank 2 As well the auto manual mode for Gadolinium control changeable by user and the current Gd load ppm in core are displayed The Manual control button for Gd addition and removal is also provided 47 FluxTC KT Nflux Nflux PDEM eur _ RD Nflux Nflux DT Perr KB For the 8 Zones ZCU _ SPD 1 Rerr G2 KluxTC KL ZCUP AVGSZCUP For the 10 Zones ZCU SPD G3 Perr G4 Avg8ZCUP 0 5 G5 AvglOZCUP ZCUP Flux Detectors Reactor LI om if Chambers Top Bottom 96 Side Side 94 Front Back UX MET 48 The rate of change in reactor power is displayed as result of the control rods movement The following time trends are displayed Reactor power thermal power and turbine power Coolant AT ROH temperature temperature Deg Actual and demanded SP Flux tilt error top bottom side side front back ZCU s and MCA average position in core Core reactivity change AK mk 49 3 9 ACR trip parameters REACTOR TRIP PARAM EM M e e ee This screen displays the parameters that cause REACTOR TRIP REACTOR STEPBACK and REACTOR SETBACK gt gt Reactor stepback is the reduction of reactor power in large step in response to certain process parameters exceeding alarm limits as a measure in
39. to remove residual heat in the fuel at the end 68 of blowdown and decay heat produced thereafter leads to the requirement for an emergency core cooling system ECC The emergency core cooling system is designed to supply emergency coolant to the reactor in two stages During the high pressure stage water is injected into the reactor core via the Emergency Core Injection ECI system on detecting a LOCA The system consists of two ECI water accumulators TK2 TK3 each accumulator is pressurized during normal reactor operation by compressed nitrogen gas A floating ball seal is located in each of the ECI accumulators At the end of injection when the water level nears the bottom of the accumulator pressure forces the ball against a seat at the bottom of the accumulator creating a seal and terminating injection This provides a passive means of defense against injection of nitrogen gas into the Heat Transport System HTS Each of the two ECI accumulators is connected to one of the two heat transport system reactor inlet headers RIH s by an injection line via MV1 and MV2 respectively One way rupture discs in the injection lines isolate the ECI system from the Heat Transport System HTS The one way rupture discs withstand the high differential pressure that is normally present in the reverse direction ECI system to HTS During normal operation the ECI system is poised to detect any LOCA that results in a depletion of HTS inventor
40. 0 full power and run the simulator gt Goto control rods amp SD rods screen record the average position of the ZCU rods and MCA rods Gadolinium load ppm Observe any flux tilt in the flux map gt Go to reactor power control Screen record the flux tilt error Record the reactivity feedback effects due to Xenon mk gt Go to turbine generator screen record the position of the main steam stop valve turbine governor control valves turbine bypass valves CSDV ASDV SG safety MSSVs gt Record the coolant pressure SG pressure and generator output 80 Press the turbine trip button on the left hand bottom corner of the screen and confirm turbine trip e Record the position of the main steam stop valve turbine governor control valves turbine bypass valves CSDV ASDV SG MSSVs e Record the reactor power SG pressure and generator output as the transient evolves e What is reactor power when turbine speed settles at 5 rpm e What is the steam flow through the bypass valve on the turbine generator screen e What is the peak SG pressure during the transient e Go to control rods amp SD rods screen record the position of the ZCU rods and MCA rods How much have the ZCU rods moved average position How much have the MCA rods moved average position Observe any flux tilt in the flux map e Go to reactor power control screen record any flux tilt error coolant pres
41. 2 reduction The basic function of the containment system is to form a continuous pressure retaining envelope around the reactor core and the heat transport system Following an accident the containment system limits release of resultant radioactive material to the external environment The containment system includes the steel lined pre stressed concrete reactor building containment structure main and auxiliary airlocks building air coolers for pressure reduction and a containment isolation system consisting of valves or dampers in the ventilation ducts and certain process lines penetrating the containment envelope This containment design ensures a low leakage rate while at the same time providing a pressure retaining boundary for LOCAs The containment system automatically closes all penetrations open to the reactor building atmosphere when an increase in containment pressure or radioactivity level is detected Measurements of containment pressure and radioactivity are triplicated and the system is actuated using two out of three logic Heat removal from the containment atmosphere after an accident is provided by local air coolers suitably distributed in various compartments inside the reactor building Hydrogen control is provided in the reactor building by passive autocatalytic recombiners that limit hydrogen content to below the acceptable limits within any significant enclosed compartment of the containment following an accident 21
42. 52 PM 2 2 54 52 PM ROH 1 ROH 2 Pressure Feed Bleed Flow a 1 13000 0 100 2 12000 0 1100 0 7 A Fw 10000 0 152 35 2 54 52 PM 2 52 35 PM 2 54 52 PM RIH 1 RIH 2 Pressure Reactor Power RIH 1 15000 REO 12500 HTS process parameters at Pump Discharge 269 6 Ces ER FL Resolution Time Scroll mm 1 Max Out Max In FLOW DIAGRAM Topic for discussion 1 Discuss the role of the steam generators as heat sink If that heat sink is lost like in this case what should be the appropriate design to back up the steam generators heat sink 87 5 2 Steam generator 1 steam flow FT irrational This malfunction causes steam flow transmitter for steam generator 1 to fail low The consequence is that the steam generator level control system for SG 1 is fooled into thinking that the steam flow from SG 1 is rapidly decreasing hence feedwater flow into SG 1 will be cutback immediately to match with false steam flow reduction in an attempt to maintain the SG level at its setpoint value In reality the steam flow from SG 1 remains at 100 nominal flow rate Because the feedwater flow is reduced to zero by the control action of the SG level control system SGLC the consequence is a rapid drop in SG 1 level When this malfunction transient occurs to reactor coolant system screen observe the steam flow from SG 1
43. 75 However as a result of turbine Bypass CSDV valves failing closed the SG pressure will increase rapidly causing further reactor power setback on high main steam pressure On the steam side ASDVs will open rapidly As the ASDVs capacity is not sufficient to relieve rising main steam pressure the main steam safety relief valves MSSVs will open to relieve rising main steam pressure that has exceeded the MSSV s lift setpoint The MSSVs will close on decreasing main steam pressure as the reactor power is being setbacked by high main steam pressure and the transient stabilizes When this malfunction transient occurs gt Go to turbine generator screen trip the turbine using the pop up control at the bottom left of the screen gt Observe that turbine bypass valves CSDV remain closed Observe the response of ASDV gt Observe that Reactor power is setbacked Can you explain the setback parameters in this incident Observe the main steam pressure transient At what pressure does the first MSSV begin to open What is the peak main steam pressure At what pressure will the MSSV start closing Explain why MSSV closing and opening again as seen in the trend At what pressure will all the MSSV be closed Observe when Reactor Setback is finished and at what main steam pressure What is the reactor power at that time gt Record and explain the transients in coolant temperature and pressure 94 Lech E
44. Advanced CANDU Reactor ACR 700 Simulator User Manual By Cassiopeia Technologies Inc Canada October 2011 FOREWORD Given the renewed worldwide interest in nuclear technology there has been a growing demand for qualified nuclear professionals which in turn has resulted in the creation of new nuclear science and technology education programs and in the growth of existing ones Of course this increase in the number of students pursuing nuclear degrees has also contributed to a large need for qualified faculty and for comprehensive and up to date curricula The International Atomic Energy Agency IAEA has established a programme in PC based Nuclear Power Plant NPP simulators to assist Member States in their education and training endeavors The objective of this programme is to provide for a variety of nuclear reactor types insight and practice in their operational characteristics and their response to perturbations and accident situations To achieve this the IAEA arranges for the supply or development of simulation programs and their associated training materials sponsors training courses and workshops and distributes documentation and computer programs The simulators operate on personal computers and are provided for a broad audience of technical and non technical personnel as an introductory educational tool The preferred audience however are faculty members interested in developing nuclear engineering courses with the suppor
45. BUNDLE CANFLEX BUNDLE 37 Elements 43 Elements CANDU 6 Fuel Channel ACR Fuel Channel Passive safety features draw from those of the existing CANDU plants eg the two independent shutdown systems and other passive features are added to strengthen the safety of the plant e g a gravity supply of emergency feedwater to the steam generators The passive safety features for ACR 700 can be summarized as follows Two independent SD systems located in low pressure and temperature moderator Low pressure and low temperature moderator surrounding fuel channels provide additional passive heat sink in the unlikely event that both the primary coolant and emergency cooling systems were unavailable Water filled shield tank surrounding Calandria would contain and maintain a collapsed core in a cooled state should moderator cooling be impaired Emergency Core Cooling ECC uses a burst disc system which functions automatically when primary system pressure drops below a prescribed level Gravity supply of Emergency feedwater to SG s Steel lined pre stressed concrete containment structure forms a safe pressure retaining envelope boundary in the unlikely event of an accident Heat removal of containment atmosphere after an accident is provided by local air coolers Hydrogen control is provided by passive autocatalytic recombiners The detailed differences between CANDU 6 and 700 are highlighted in the following
46. C HX1 Avg Coolant T F 690 LTC sump1 F 690 f LTC sump 2 LTC Mv4 L Emergency Core Cooling Flow Diagram Emergenc Reactor Reactor Generator Output Primary Coolant Core Main STM Press Neutron Pwr 9 Thermal Pwr Average ROH Pressure Flow kg s gop STM Flow Freeze Iterate Core Cooling kPa SE CC 119 6 REFERENCES 1 AECL Document ACR 700 Technical Description 10810 01371 TED 001 Revision 1 March 2004 2 AECL Document PRELIMINARY DESIGN DESCRIPTION OF PRESSURE AND INVENTORY CONTROL SYSTEM ACR 700 10810 33310 200 001 Revision 1 Feb 2004 120
47. CTOR POWER SETPOINT RATE limited by the maximum rate of 0 8 of full power per second DEMANDED POWER SETPOINT is the incremental power target which is set equal to current reactor power 96 rate s program cycle time sec In this way the DEMANDED POWER STEPOINT is ramping towards the REACTOR POWER SETPOINT target at the accepted rate of change From the DEMANDED POWER SETPOINT CURRENT REACTOR POWER TARGET RATE CURRENT RATE OF CHANGE OF REACTOR the POWER ERROR can be determined as follows Nflux PDEM KRATE NfluxP _RD Nflux Nflux DT Perr KB Where Perr Reactor Power Error Nflux Current Bulk Reactor Power NfluxP Current Bulk Reactor Power in previous RRS program cycle PDEM Demanded Power Setpoint RD Reactor power rate demanded sec DT Time sec between successive execution of the RRS program KB Gain constant for difference between current power versus demanded power KRATE Gain constant for the rate difference between the current reactor power rate versus the demanded rate gt The FLUX TILT component of the reactor consisting of multi cells representing respective section of the core can be determined as follows 45 FluxTC KT Nflux Nflux where Flux TC Flux tilt component for i cell in the reactor core Nflux average reactor flux in core Nflux reactor flux at the i cell of the reactor core Havi
48. Ee dE DEENEN EE bete EE Ce 96 5 10 PRZR HEATERS 2 TO 6 TURNED ON BY MALPUNGCTION cene eene nna 96 5 11 RC INVENTORY FEED VALVE CV 12 FAILS OPEN 98 5 12 RC INVENTORY BLEED VALVE CV5 FAILS OPEN eren en nennen erre inneren rennen ne 100 5 13 PRZR PRESSURE RELIEF VALVE CV22 FAILS OPEN 101 5 14 ONE BANK OF MCA RODS DROPS eese entere enne te than nisse tentare nasse esee tates asas es eese terna nne enn 103 5 15 ALL MCA RODS STUCK 105 5 16 REACTOR SETBACK STEPBACK BOTH PAI 106 5 17 LOSS OF ONE RC PUMP P1 5 18 0866 PUMPS a reete eee ete EE 111 5 20 PRIMARY COOLANT RIH 1 LOCA BREAK 1 INTRODUCTION 1 1 Purpose This publication consists of course material for workshops on the advanced heavy water reactor known as the Advanced CANDU Reactor ACR 700 MWe simulator Participants in the workshops are provided with instruction and practice in using the simulator thus gaining insight and understanding of the design and operational characteristics of ACR 700 nuclear power plant systems in normal and accident situations This manual is written with the assumption that the readers already have some knowledge of the CANDU and ACR Therefore no attempt has been made to provide detailed descriptions of each individual ACR subsystem which should be available in publications from Atomic Energy of Canada Ltd AECL
49. Gen Level Hi Loss RC Pmp s Reactor Pwr amp Thermal Pwr PLANT MODE REACTOR LEADING o 110 0 POWER RATE amp TARGET LOAD CURRENT OPERATOR CONTROLLED VARIABLE TARGET INPUT TARGET RANGE TARGET LOAD D 0 00 100 00 5TO 100 0 17 54PM 2 2 20 11 Cal om Current Target Load amp Turbine Pwr STEAM GENERATOR PRESSURE SETPOINT CONTROL dar vaina 244 keA sPMopE oo 6400 sp kPa 11PM Resolution Time Scroll E Max In Generator Output Primary Coolant Core Main STM Press T Neutron Pwr 95 Thermal Pwr Average ROH Pressure Flow kg s BOP STM Flow kPa EH ES __ __ step back Redd ad EE ER Ir EE z No flow in path ECI Signal Initiated CONTAINMENT m Flow in path Reserved Reserved Water Tank MSSV1 MSSV2 ECI Injection Tanks 340 87 gt To Main Turbine j MSSV3 Sei Sau wi B BR RSW By m zs MV3 MSSV4 PRZR P 1628 Main Steam Header Surge MV1 LTC MV6 NOTE Pipings and valvings shown on screen only illustrate main arrangements for ECI flow paths Not all details are shown See the Emergency Core Cooling Flow Diagram for details LTC MV2 Fl 0 Break Flow 686 LT
50. H 2 coolant channels in core Pressure and inventory control systems are shown on subsequent displays The parameters displayed are gt Reactor pressure KPa at ROH 1 ROH 2 28 gt Reactor core flow kg sec Average reactor coolant temperature C gt Average fuel temperature C gt Status of the four coolant heat transport pumps HTS P 1 2 3 4 e The two steam generators are individually modeled along with balance of plant systems The parameters displayed are y y SG 1 2 level m SG 1 2 steam flow kg sec SG 1 2 steam pressure kPa SG 1 2 steam temperature C Total steam flow kg sec from the steam generators Steam flow to main steam relief valves MSSV s Opening status of the main steam relief valves MSSV s The MSSV s are represented by one valve symbol that is in the event that any MSSV opens the valve symbol colour will be red green when all MSSV s are closed Steam flow to Atmospheric Discharge Valves ASDV and Condenser Steam Discharge Valves CSDV The ASDV and CSDV are respectively represented by one valve symbol that is in the event that any valve opens the valve symbol colour will be red green when all valves are closed e Steam Turbine is a simple model The parameters displayed Status of turbine control valves is indicated by their colour green is closed red is open gt Moisture separator and reheater MSR drains flow kg sec
51. IP status is shown as NO green or YES yellow the trip be reset here note that SDS RESET must also be activated before reactor shutdown system will begin withdrawing the Shutdown Rods The REACTIVITY CHANGE mk of each device and parameter from the initial 100 full power steady state is shown These include 1 SHUTDOWN RODS SOR 60 MK for full SOR drop 2 ZONE CONTROL UNITS ZCU 18 ZCUs total reactivity worth is 9 MK 4 5 MK fully withdrawn 4 5 MK fully inserted in core The maximum speed of ZCU movement gives or 0 2 mk per sec of reactivity rate change 39 gt Note control buttons are provided on the Reactivity pop up to allow users to increase and decrease the reactivity worth of the ZCU online This is to facilitate user s observation of the ZCU control system response with various design values of total MK worth as an educational exercise MECHANICAL ABSORBER ROD UNITS MCA total of 8 absorber rods 4 in the core center region 4 in the core outer region Total reactivity is 12 MK for full insertion The speed for all the absorber rods is constant and the full insertion travel time is 120 sec For the simulator the absorber rods are divided into 2 banks with each bank s reactivity worth of 6 MK XENON full power steady state Xenon load 26 MK peak Xenon load 12 hours after full power trip 63 MK FUEL TEMPERATURE reactivity feedback 0 014 MK deg C from 687 to 787 deg C
52. Main STM Press Iterat amp z Neutron Pwr Thermal Pwr Average ROH Pressure Flow kg s BOP STM Flow erate Pressurizer kPa cce Reactor Tro Turone os iT TST 100 00 11997 76 711 87 100 5 13 PRZR pressure relief valve CV22 fails open This malfunction transient causes depressurization of the pressurizer with steam vapor going to the bleed condenser through the failed opened pressure relief valve CV22 As the pressure is decreasing in the pressurizer the electric heaters will be turned on As well pressurizer level will rise with decreasing pressure The rising pressurizer level will cause the bleed flow to increase trying to reduce coolant inventory in the pressurizer Although the electric heaters are turned on they cannot cope with the pressure loss caused by the failed CV22 venting to the bleed condenser As a result the coolant pressure keeps dropping during this transient leading to reactor trip by low reactor outlet header pressure When this malfunction transient occurs to coolant inventory and pressurizer screen Observe that CV22 fails open Observe the pressurizer pressure transient and level transient Note that the electric heaters will turn on Record the bleed flow to the letdown condenser Continue to monitor coolant pressure record when reactor trip occurs Monitor flow through CV22 as the pressurizer pressure continues to decreas
53. Outlet K Header 2 Pressure Flux Tilt Top bottom Side side Front back ZCU 8 zones group control rods average position ZCU 10 zones group control rods average position MCA Bank 1 rods average position C C 76 MCA Bank 2 rods average position Gadolinium Concentration Pressurizer Temperature Coolant Feed Flow Kgs S Coolant Bleed Kg s Flow Main Steam Pressure Total steam flow Kg s from steam generators Total Feedwater Kg s Flow e Explain the main changes main steam header pressure rises first then drops back to the steam pressure setpoint value although the steam pressure setpoint value is unchanged gt Why steam generator s level drops initially and then recovers Turbine power lags target load but follows it nicely However the reactor neutron amp thermal power overshoot below 90 power but recover later But their values drift up and down for sometime before they stabilize Recall previous power maneuvering in reactor leading mode the reactor neutron amp thermal power decrease orderly and do not drift as much during power changes Can you explain why this occurs in this power maneuvering in turbine lead mode What is the difference in the way reactor power is controlled in reactor lead mode versus turbine lead mode e Continuing the above operation raise UNIT POWER to 100
54. The main steam lines supply steam from the two steam generators in the reactor building to the turbine through the steam balance header in the turbine building at a constant pressure The system controls the steam generator pressure using the condenser steam discharge valves CSDVs and the atmospheric steam discharge valves ASDVs Main steam safety valves MSSVs are provided for overpressure protection of the steam generator secondary side The feedwater system takes hot pressurized feedwater from the feedwater train in the turbine building and discharges the feedwater into the preheater section of the steam generators Main steam isolation valves MSIVs are provided to isolate the main steam supply to the turbine in the event of steam generator tube leak after reactor shutdown when the long term cooling system is placed in service and the heat transport system is depressurized The feedwater system controls the feedwater flow to maintain the required steam generator level The flow diagram for the system is shown below 14 Feedwater System Flow Diagram 15 2 6 Balance of Plant The balance of plant BOP consists of the turbine building steam turbine generator and condenser the feedwater heating system with associated auxiliary and electrical equipment The BOP also includes the water treatment facilities auxiliary steam facilities pumphouses and or cooling towers main switchyard and associated equipment to provide all conv
55. activity available for the ZCU rods even if they are fully withdrawn their combined reactivity is insufficient to compensate for the negative reactivity imparted from dropping the bank of MCA rods into core As a result the reactor power is decreasing coolant pressure is decreasing As well the main steam pressure is decreasing leading to turbine runback and a subsequent turbine trip on zero forward power The transient will evolve with the reactor power slowly decreasing to zero due to Xenon buildup When this malfunction transient occurs gt Go to control rods amp SD rods screen observe that one bank of MCA Rods has been dropped into the core Record the overall reactivity change and reactor power immediately after the malfunction is initiated Go to reactor power control screen observe that the ZCU rods are withdrawing Record the reactivity mk change Goto reactor coolant system screen and observe the coolant pressure transient Go to turbine generator screen observe the main steam pressure transient Note the turbine runback is in progress Go back to control rods amp SD rods screen record the overall reactivity change again Record the reactor power Describe and explain the long term evolution of this transient 103 Turbine Trip ROH Press LoLo Step Back Redd Turbine Runback LL eege Opn Low Fwd Pur Trip Main Trip a cce eem PRZR RCTR
56. ader Note For simplicity the following systems are not modeled in the ACR 700 simulator a Moderator System b Condenser and Condensate System c Shutdown System 2 SDS2 25 3 1 3 2 Simulator startup Select program icon ACR Simulator for execution Click anywhere on simulator screen Click OK to load full power IC The simulator will display the ACR plant overview screen with all parameters initialized to 100 full power At the bottom right hand corner click on Run to start the simulator Simulator initialization If at any time it is necessary to return the simulator to one of the stored initialization points do the following 3 3 1 2 3 4 5 6 7 8 9 10 11 12 13 14 Freeze the simulator Click on IC Click on Load IC Click on FP_100 IC for 100 full power initial state Click to Load C ACR FP_100 IC Click YES to Load C ACR FP_100 IC Click Return Start the simulator operating by selecting Run List of ACR simulator display screens ACR Plant Overview ACR Control Loops ACR Control amp Shutdown Rods ACR Reactor Power Control ACR Trip Parameters ACR Reactor Coolant System ACR Coolant Inventory amp Pressurizer ACR Coolant Inventory Control ACR Coolant Pressure Control ACR Turbine Generator ACR Feedwater amp Extraction steam ACR MW demand SP amp
57. age will say MAN SP OK e The following trends are displayed Yu YY UY Reactor neutron power reactor thermal power turbine power Steam flow to deaerator Kg s Deaerator pressure KPa amp setpoint KPa Main steam header pressure KPa High pressure heaters HX5A HX5B extraction steam flows Kg s Steam generator level m 65 3 16 MW demand setpoint SP and steam generator pressure control SGPC 1 RCTR Reactor Pwr amp Thermal Pwr 110 0 PLANT MODE REACTOR LEADING a TRML 100 0 CONTROLLED VARIABLE pnm OPERATOR TARGET LOAD D 100 00 100 00 5 TO 100 STEAM GENERATOR PRESSURE SETPOINT CONTROL Woh Dir D ekpa rn DI _6 05 POWER RATE amp TARGET LOAD mS TARGET INPUT TARGET RANGE 0 0 10 01 57 PM 10 02 04 PM 0 10 0 01 TO 1 Main Steam Hdr Pressure amp SP SP P Current Target Load amp Turbine Pwr 101 04 Loap ER 90 07 SE d 30 0 70 0 60 0 50 0 40 0 30 0 20 0 10 0 0 0 10 01 57 PM 10 02 04 PM SP Recovery PRESSURE SP CHANGE RATE D Resolution Time Scroll es Max Out sse Max In MW Demand Reactor Reactor Generator Output Primary Coolant Core Main STM Press I amp SGPC Neutron Pwr Thermal Pwr Average e Pressure Flow kg s em Flow ETE terate semen I ic war Sg
58. ajor plant parameters Reactor neutron power Reactor thermal power Generator output Primary coolant average reactor outlet header ROH pressure kPa Core flow kg sec Main steam pressure KPa Balance of Plant BOP steam flow Kg sec Feedwater Flow Kg sec Average fuel temperature Deg e The bottom left hand corner allows the initiation of two major plant events gt Reactor trip gt Turbine trip these correspond to hardwired push buttons in the actual control room e The box above the Trip buttons shows the display currently selected i e plant overview by clicking and holding on the arrow in this box the titles of the other displays will be shown and a new one can be selected by highlighting it The remaining buttons in the bottom right hand corner allow control of the simulation one iteration at a time iterate the selection of initialization points IC insertion of malfunctions malf and calling up the help screen if the on line help program is provided 27 3 5 ACR plant overview Press Lo o Iurbine Runbazk Gen Ereaker Ces Lee MN TCR 38 san boR _ _ 5 _ 5 use noB mpscio spe _ ACR 700 TECHNICAL FEATURES H E s asov Oi gp Pd uL e Sg Seen Outlet P
59. am pressure has immediate impact on the heat transfer of the steam generators As a result there will be transients on coolant temperature and pressure But turbine trip will occur very quickly causing setback of reactor power and the transients in the reactor and primary coolant will stabilize When this malfunction transient occurs Go to turbine generator screen observe the turbine governor position Observe the main steam pressure transient What is the peak steam pressure Explain how the turbine bypass valves operate Observe turbine power is decreased very rapidly followed by turbine trip Repeat this malfunction again while the reactor coolant system screen is displayed Observe RIH ROH temperature transients It is necessary to change the scale of the trend accordingly in order to see the transient better What is the peak RIH ROH temperature during this malfunction Explain why the RIH ROH temperatures go up 90 bine Trin _ROHPresstoto LL stepBackRegd L SetbackRedd_ TurbineRunback 0 Gen Breakeropn _HineutPwrtogr RoHpressri vezmpen z n IL ass iR RIH 1 2 ROH 1 2 Temp P1 2 3 4 Flows to RIH S daa 2000 1000 HTS process parameters at Pump Discharge D p 56 44 PM 3 59 07 PM p 56 44 PM 3 59 07 PM 280 8 7 280 7 ROH 1 ROH 2 Pressure es Feed Bleed Fiow a rj 1787 9 d bd 1318 ng
60. an advanced ACR The passive emergency core cooling system requires no operator actions to mitigate design basis events like loss of coolant accident LOCA The system relies on natural forces such as gravity natural circulation and compressed gas Only few valves are used in the system supported by reliable power sources The ECC function is accomplished by two sub systems The emergency coolant injection System for high pressure coolant injection after a LOCA The Long Term Cooling LTC system for long term recirculation recovery after a LOCA The LTC system is also used for long term cooling of the reactor after shutdown following other accidents and transients Following a loss of coolant accident the reactor shutdown and emergency core cooling systems acting together must as a design target prevent excessive fuel damage In the event of a major break in the heat transport system the water escapes through the break depressurizing the system the blowdown phase The reactor is tripped automatically The combination of increase in pressure differential across the fuel sheath caused by the gaseous fission products and the increase in sheath temperature is a factor affecting the sheath failure threshold during blowdown If the threshold is exceeded the sheath can swell and could result in sheath rupture However during blowdown the sheath temperature increase is limited and excessive sheath failures are prevented The need
61. and design review by AECL 52 3 10 reactor coolant system Eee 5 HEADER gt e 86205 M e 6400 l F 75845 m HTS process parameters at Pump Discharge 1 2898 ur HEAR deel y H E Ei o J ees gn P1 D e r A ARI ON Ju 1 RIMAT SS e Time Scroll D Row oras DIAGRAM Main STM Press TM Generator Output Primary Coolant PR P 96 ce owns et Pressure Fw 6 o s BOP STM Flow 00 Freeze iterate FW Flow Reactor Reactor Trp Turbine TT Fuel Temp Em EJ This screen shows a layout of the Heat Transport System HTS two steam generators four heat transport pumps reactor inlet header RIH 1 2 reactor outlet header ROH 1 2 reactor vessel with coolant feeder piping The primary coolant is circulated through four heat transport pumps into the core through the through two reactor inlet headers known as RIH 1 and RIH 2 respectiively After entering the RIH 1 2 the coolant then travels through the fuel channels in the core and exits the core at two reactor outlet headers known as ROH 1 2 The two ROHs are connected to two steam generators respectively The heated coolant then flows down through the two steam generators where the heat is transferred to the secondary system The primary coolant is then taken from the b
62. and each heater may be selected to START STOP or RESET gt in order to control the variable Heater 1 MANUALLY one must use the pop up menu to switch the control mode from AUTO to MANUAL first then the control signal to the Heater 1 will be frozen as shown in the numeric value display Observe the display message above the Heater control If it says MAN O P that means Heater 1 can now be controlled by the MAN pop up menu If it says MAN O P NOT that means the MANUAL control signal from the MAN pop up and the frozen control signal to the Heater does not match One must then use the MAN pop up menu to enter a value equal to the frozen numeric value display then the message will say MAN O P e In the next section PRESSURIZER RELIEF VALVES CONTROL is via CV22 and CV23 These are normally in AUTO mode but may be placed on MANUAL and the valve opening can be controlled manually via pop up menus 60 In the third section the PRESSURIZER SPRAY VALVES CONTROL is via SCV1 and SCV2 These are normally in AUTO mode but may be placed on MANUAL and the valve opening can be controlled manually via pop up menus NOTE in order to control the pressurizer relief valves or pressurizer spray valves MANUALLY one must use the pop up menu to switch the control mode from AUTO to MANUAL first then the control signal to the control valve will be frozen as shown in
63. at a rate of 0 3 FP sec e When reactor power has returned to 100 and the parameters have stabilized freeze the simulator and record the parameter values in column 3 100 stabilized of Table III Go to reactor power control screen and record parameters in column 3 of Table IV Note any major difference in parameter values between column 3 and column 1 Can you explain why the differences in parameter values if any 77 4 3 Power level reduction to 0 Initialize the simulator to 100 FP using reactor lead mode reduce reactor power in 25 steps at 0 5 sec During power changes go to the following screens and record the parameters in Table V y Control rods amp SD rods Reactor power control Reactor coolant system ACR coolant inventory amp pressurizer Turbine generator Feedwater amp extraction steam Under comments please note type of parameter change as a function of reactor power 0 100 FP constant linear increase or decrease non linear increase or decrease Note any alarms encountered during the power changes In case reactor setback or stepback occurs the trip parameters screen will indicate the causes for such alarms 78 TABLE V POWER LEVEL REDUCTION TO 0 eme ep sm ase commen EE ZCU 8 zones group rods ezes EES ZCU 10 zones group rods _ pop o MCA Bank 1 Rods MCA Bank 2 Rods p eng pasio
64. ater tank to the steam generators e A separate secondary control area is provided as a backup to the main control room for certain emergency conditions e Distributed control systems control the plant routinely freeing the operator from mundane tasks thus reducing the likelihood of operator error The safety system responses are automated to the extent that no operator action is needed for a minimum of eight hours following most design basis accidents The safety systems are those systems designed to quickly shut down the reactor remove decay heat and limit the radioactivity release subsequent to the failure of normally operating process systems These are the shutdown system number 1 SDS1 shutdown system number 2 SDS2 emergency core cooling ECC system and containment system The safety support systems are those that provide services needed for proper operation of the safety systems e g electrical power cooling water instrument air a Shutdown System No 1 SDS1 SDS1 quickly terminates reactor power operation and brings the reactor into a safe shutdown condition by inserting shutoff rods into the reactor core Reactor operation is terminated when a certain neutronic or process parameter enters an unacceptable range The measurement of each parameter is triplicated and the system is initiated when any two out of the three trip channels are tripped by any parameter or combination of parameters b Shutdown S
65. croll AVG FUEL 714 70 4 H gt s24 Max Out Qe In Reactor Coolant Reactor Reactor Generator Output Primary Coolant Core System Neutron Pwr Thermal Pwr Average Mu ee Flow kg s EES DEER iz 7135 17 97 5 11 RC inventory feed valve CV12 fails open This malfunction causes the reactor coolant feed flow to reach the maximum The immediate impact to the reactor coolant system is increased coolant inventory in the system As a result the pressurizer level will increase leading to increase in pressurizer pressure This is due to the fact that the vapor space in the pressurizer has been reduced by higher liquid mass in the pressurizer because of increased inventory The increased pressurizer pressure is offset by the spray action which comes into effect on high pressurizer pressure But the spray will further increase the pressurizer level The high pressurizer level will cause the inventory control system to increase the bleed flow by opening the Bleed Valve CV5 As a result the bleed condenser level will increase Overtime the coolant feed flow and the coolant bleed flow will balance out and the transient will stabilize When this malfunction transient occurs the coolant inventory and pressurizer screen observe that CV12 is 100 open and record the feed charging flow kg s Observe the coolant pressure transie
66. ction between all simulated systems e Overall unit power control with reactor leading mode or turbine leading mode e Unit annunciation time trends e Computer control of all major system functions e Emergency Core Cooling System ECC Simple Model containment amp for DISPLAY OPERATOR CONTROLS e ACR feedwater extraction steam e Feed pump on off e all amp operation e SG level controller mode le one Auto or manual e level setpoint during operation e level control valve e all opening during manual operation e extraction valves opening control e one changes MALFUNCTIONS level control isolation valves fail closed level control valve fails open level control valve fails closed Auto e main pump trips feedwater main steam safety relief valves MSSV open steam e steam header break e steam flow transmitter failure eACR turbine generator e turbine trip e turbine run back synchronization condenser discharge CSDV e atmospheric steam discharge valve ASDV e ACR Plante reactor power Overview setpoint and rate e ACR Control entry in reactor lead Loops mode e ACR turbine load Demand SP amp setpoint MW and SGPC loading rate entry in turbine lead mode valves e ACR Passive Core Cooling break e turbine spurious trip condenser e turbine run up and discharge CSDV steam closed steam valves failed e reactor inlet he
67. duce only modest reactivity changes e The control and shutdown devices in the low pressure moderator and are not subject to large hydraulic forces The equilibrium core has a significantly negative power reactivity coefficient which provides inherent protection against transients with inadvertent increase of reactor power e The void reactivity coefficient is small and negative and offers a good balance of inherent nuclear protection between loss of coolant accidents LOCA and accidents with fast cooldown of the heat transport system e Natural coolant circulation can remove decay heat from the fuel if Class IV electrical power to the heat transport pump motors is lost e Two independent shutdown systems are provided Each system can shut down the reactor for the entire spectrum of design basis events e Emergency core cooling ECC is provided by an emergency coolant injection system which injects water into the heat transport system after a LOCA A long term cooling LTC system provides adequate decay heat removal from the reactor core in the recovery recirculation phase after a LOCA e For a loss of the main feedwater pumps and or Class IV electrical power the auxiliary feedwater pumps with power supplied from the Class III power systems provide effective cooling with the reactor shut down The auxiliary feedwater supply is also backed up by 16 passive emergency feedwater with gravity water supply from the reserve w
68. e At what pressure will the flow from CV22 stop Why ee ee ee ee ae SCHT PRZR Spray Flow Coolant Feed amp Bleed Flows BLEED CDSR Pressure Resolution Time Scroll ROH PRESSURE CONTROL MODE A DI Norma NOTE Screen only shows typical arrangements Max Ou Max In Not all valves are shown final design may change Primary Coolant Main STM Press I Average ROH Pressure BOP STM Flow terate NETTES rure BEEN Inventory Core amp Pr a Flow kg s Reactor Trip Turbine Trip Reactor Reactor Generator Output Neutron Pwr 96 Thermal Pwr 101 TO MAIN STEAM HEADER e reel T 2777 m D 2 HTS process parameters at Pump Discharge FJ P 10781 Gre e LII Wa ST actor Ri Neutron Pwr 96 Thermal Pwr Average ROH Pressure Flow kg s kPa eactor Generator Output Primary Coolant Core 102 5 14 One bank of MCA rods drops This malfunction event will drop one bank of absorber MCA rods into core imparting large negative reactivity into the core This leads to large reduction of reactor power The reactor regulating system RRS will immediately withdraw the ZCU rods for reactivity compensation However because there is only limited re
69. e This is due to the fact that the vapor space in the pressurizer has been increased by reduced liquid mass in the pressurizer because of decreased inventory The decreased pressurizer pressure will turn on the pressurizer heaters The low pressurizer level will cause the inventory control system to increase the feed flow by opening the feed valve CV12 Over time the coolant bleed flow and the coolant feed flow will balance out and the transient will stabilize When this malfunction transient occurs Go to the ACR coolant inventory and pressurizer screen observe that CV5 is 100 open and record the bleed flow kg s Observe the coolant pressure transient and that the pressurizer heaters turn on Observe the pressurizer level and record the feed flow kg s Observe the bleed condenser level E E EE Hi Neut Pwr LogR ROH Press Hi Main Stm Pres Hi Stm Gen Level Hi Loss RC Pmp s scvi PRZR RCTR OUT Pressure BLEED CDSR Level PRZR Spray Flow 10 11 07 10 11 07 Coolant Feed amp Bleed Flows BLEED CDSR Pressure D makeup 10 08 48 PM 10 11 07 0 Resolution Time Scroll FLOW DIAGRAM ROH PRESSURE CONTROL MODE DI NoRMAL NOTE Screen only shows typical arrangem tits Not all valves are shown final design may change Max Out In Inventory Reactor Reactor Generator Output Coolant Core
70. e heat transport system also provides a barrier to the release of radioactive fission products during normal operation to ensure that radiation doses to plant staff remain within acceptable limits It is designed to retain its integrity under normal and abnormal operating conditions 2 4 Moderator System Valve Normally Open o Check Valve Cover Gas System Recombination Units Reactor Head Tank Compressors 020 Sampling Circulation Pump d il v Purification System 12 RCW Neutrons produced by nuclear fission are moderated by the heavy water in the calandria The heavy water moderator is circulated by the moderator pumps through the calandria at a relatively low temperature and low pressure and cooled by the moderator heat exchangers The moderator heat exchangers remove the nuclear heat generated in the moderator and the heat transferred to the moderator from the fuel channels Helium is used as a cover gas over the heavy water in the calandria Chemistry control of the moderator water is maintained by the moderator purification system The moderator system also acts as a back up heat sink under certain postulated accident conditions 13 2 5 Steam and Feedwater System CJ EMERGENCY KOMMEN Ch EMERGENCY AR ME PESTER FEPDIATER ET BLONCORN KE Yr 1 STEAM 4 STEAM GENERATOR 1 GENERATOR 2 W i Wei FEEDVATER Main Steam System Flow Diagram
71. e the core to part way or all the way to their fully inserted position at core 4 Excess reactivity due to fresh fuel and decay of xenon following a long shutdown are compensated by adding poison boron or Gadolinium to the moderator Note only Gadolinium poison is modeled in this simulator 5 Reactivity variations due to on power refuelling during equilibrium operation since the reactor is fuelled continuously and on power at a rate which keeps the reactor critical the control requirements for refuelling are within the range of the zone controller response Soluble poison concentration is normally near zero For a standard 2 bundle shift fuelling scheme the reactivity increase due to refuelling in an average channel is less than 0 2 mk This reactivity change is sufficiently controlled by the zone controllers Note on power refuelling is not modeled hence the reactivity variations due to on power refuelling will not be observed in this simulator 6 Rapid shutdown of the reactor is by dropping solid control absorbers shutdown rods into the core and or by the fast injection of large amounts of liquid poison into the moderator Note SDS1 is modeled in this simulator Specific details regarding the respective reactivity devices are provided below a Zone Control Units ZCU The zone control system consists of nine vertical assemblies with two independently moveable segments in each assembly hence 18 ZCUs Reactivity is
72. edwater and extraction steam Neut Thrm Pwr Turb Pwr D A stm flows Extraction Main Main Stm RES 000 Pressure Aso d og To HP heaters BOILER LEVEL SP comp 5 D peel sc Lv se v 13 5000 5 Ext Stm Flows e SG Level 5 0 01 57 Resolution 4 m m 1 J FLOW DIAGRAM Max Out Max In NOTE Screen only illustrates main feedwarer amp extraction steam flow arrangements Not all valves are shown See Flow Diagram for details Reactor Reactor Generator Output Primary Coolant Core Main STM Press Neutron Pwr Thermal Pwr Average ROH Pressure Flow kg s Iterate KPa ETH This screen shows the portion of the feedwater system that includes the condenser low pressure heater deaerator the boiler feed pumps the high pressure heaters and associated valves with the feedwater going to the steam generator level control valves after leaving the HP heaters The following display parameters and pop up controls are provided e Main steam header pressure KPa steam flow through the turbine governor valve and the bypass valve Kg s e Deaerator level m and deaerator pressure KPa extraction steam motorized valve status and controls from turbine extraction as well pressure controller controls for main steam extraction to deaerato
73. em ILL S 12 32 42 PM 12 32 49 PM Reactor Coolant Pressure Main Stm Pressure 13000 0 10000 0 8000 0 6000 0 4000 0 2000 0 Electrical Out Setpoint Turbine Leading Click on respective box for further details of the specific Control Loop CONTROL Generator Output Primary Coolant Core Average ROH Pressure Flow kg s BOP STM Flow E Fwriow 2 Main STM Press Freeze Iterate Reactor Reactor Neutron Pwr Thermal Pwr Reactor Trip Turbine Trip The plant power control function of a ACR type NPP is performed by two separate control modes one for the turbine generator called turbine leading and the other one for the reactor called reactor leading These two distinct modes of overall plant control can be switched between each other and are well coordinated for plant startup shutdown power operations of all kinds and for plant upset conditions In the turbine leading control mode generator power is controlled according to the power demanded by means of a remote reference value e g operator input and or by a value derived from the actual generator frequency deviation from the grid Using this deviation from setpoint the reactor power is adjusted using the main steam pressure error i e deviation from normal
74. em relief valve fails two steam generators System HTS pumps open six equivalent lumped e ACR Coolant e coolant makeup ecoolant feed valve reactor coolant channels Inventory amp pumps fails open e Fuel and coolant heat Pressurizer pressurizer ecoolant bleed valve transfer simulated for 18 e ACR Inventory pressure control fails open reactor zones Control heaters spray epressurizer heaters e Pressure and inventory e ACR Pressure pressure control 2 to 6 turned control which includes Control valve relief valve ON by pressurizer bleed epressurizer level malfunction condenser feed amp bleed control reactor inlet header control and pressure regulating coolant break relief coolant makeup feed amp bleed flow eloss of one HTS e Operating range is from via control valves pump zero power hot to full eisolation valvesjeLoss of two HTS power for coolant feed pumps in one loop and bleed 24 SIMULATION SCOPE e SG dynamics including shrink and swell effects e Steam supply to turbine and reheater e Turbine by pass condenser e Extraction steam to feed heating e Steam generator pressure control eSteam generator level control SG feed system to e Simple turbine model e Mechanical power and generator output proportional to steam flow e Speeder gear and governor valve allow synchronized and non synchronized operation e Turbine steam bypass e Fully dynamic intera
75. entional services to the ACR 700 two unit plant Two steam generators are provided in the heat transport system They discharge steam into a common header located in the turbine building that supplies the required steam to the turbine generator and the auxiliary steam systems The power generating equipment consists of the following e turbine generator set with a nominal gross output of 753 MW e This consists of a tandem compound reheat condensing type steam driven single shaft turbine composed of one high pressure and two low pressure cylinders with a thermal cycle involving two stage moisture separator reheater vessels located between the high pressure turbine exhaust and the low pressure turbine inlets The generator is cooled with water and hydrogen and provided with a static excitation system e condenser with tubes at right angles to the turbine axis e regenerative feedwater heating system with low pressure stages deaerating feedwater heater and high pressure stages e Other auxiliaries associated with the turbine generator set 2 7 Safety Systems The ACR safety design has the following inherent and engineered safety characteristics e On power refuelling assures that very little reactivity needs to be held up in movable control devices or in neutron absorbent material dissolved in the moderator no chemicals are added to the reactor coolant for reactivity control Thus any malfunctions in the reactor control system pro
76. ew values for pressure SP TARGET in MPa and the pressure SP change rate in MPa minute Observe that the SP value changes immediately after the new SP target and rate are accepted As well the main steam header pressure shown in the display will be changed At any time if one wants to return the original pressure setpoint just press the button SP recovery once It can observe that the pressure SP will recover to 6400 KPa and the main steam header pressure will follow accordingly The following trends are provided gt Reactor neutron power reactor thermal power Main steam header pressure KPa amp setpoint KPa Current target load and turbine power Steam generator 1 amp 2 level m 67 3 17 ACR passive cooling CONTAINMENT m Flow in path MSSV3 Main Header LTC MV3 ROH interconnect Avg Coolant T 300 58 Diagram for details FB 0 Generator Ven H Primary Coolant Main STM Press EYE Neutron Pwr Thermal Pwr Pressure Flow or BOP STM Flow EN FW Flow REH NENNEN S Fuel Temp Reactor Trip Turbine Trip NOTE Pipings and valvings shown on screen only illustrate main arrangements for ECI flow paths Not all details are shown See the Emergency Core Cooling Flow This screen shows the passive core cooling system in
77. fully open thus allowing the pressurizer to assist in maintaining coolant pressure and mass inventory at setpoint The following time trends are displayed Reactor neutron power 96 reactor thermal power 96 Reactor outlet pressure KPa amp setpoint KPa Pressurizer level m amp setpoint m Pressurizer relief valve position 61 3 14 turbine generator Main Steam Header P To HP heaters STATION 110 0 6399 4 services 50 00 e MSV Inlet GENERATOR OUTPUT J MW RCTR __RCTRNeut Thrm Pwr Pwr Generator Outp MW E PM um EE ee Fa 10 01 57 PM 10 02 04 Turb Steam CSDV ASDV MSSV Flow 1200 0 m MSV Inlet Pressure MAN SP NOT OK MSSV1 TURBINE TRIP STATUS Combined d MSSV2 wears fi k Cast RESET C Speeder Gear 101 88 Deg SPDR GEAR CONTROL auo H TURBINE Resolution BE RUNBACK D 4 i Flow pracram DIAGRAM Max Out Max In NOTE Screen only illustrates main steam arrangements Not all valves are shown See Flow Disaram for details Reactor Reactor Generator Output Primary Coolant Core Main STM Press Neutron Pwr Thermal Pwr Average ROH Pressure Flow kg s BOP STM Flow ROSE FW Flow Lg WEE EES EICH BEER EENS This screen shows the main parameters and controls as
78. g parameters in the full power column 1 before power maneuvering Go back to MW demand setpoint amp SGPC screen Reduce unit power in the turbine lead mode i e gt Select the plant mode to be turbine lead gt Select TARGET LOAD pop up menu npop up menu lower target to 90 00 at a rate of 0 3 sec gt accept and return Observe the response of the displayed parameters until the transients in reactor power and steam pressure are completed without freezing the simulator and or stopping labview When the parameters have stabilized freeze the simulator and record the parameter values in column 2 90 stabilized of Table III Go to reactor power control screen and record parameter values in column 2 of Table IV TABLE III PLANT LOAD MANEUVERING TURBINE LEAD 1 Parameter Unit 2 3 Comments 90 96 return to stabilized 100 stabilized Reactor Neutron Power Reactor Thermal Power Main Steam Header KPa Pressure Main Steam Pressure KPa Setpoint 75 TABLE IV PLANT LOAD MANEUVERING TURBINE LEAD 2 Comments return to 100 stabilized stabilized Average Coolant Temperature from ACR Reactor Coolant Screen Average Core Flow Average Fuel Temp Coolant Delta T ROH temp RIH temp Reactor Outlet Header 1 Pressure Reactor Inlet Header 1 Pressure Reactor Inlet Header 2 Pressure Pressurizer Level M Reactor
79. ge 7 7279 9 rmm E Ee Fei aer 11 50 50 11 53 09 11 50 50 AM 11 53 09 ROH 1 ROH 2 Pressure Feed Bleed Flow ei 11 50 50 AM 11 53 09 11 50 50 AM 11 53 09 RIH 1 RIH 2 Pressure Reactor Power RIH 1 14000 110 12000 75 po Em x CH Max Out Max In Core Main STM Press ka s BOP STM Flow REES Primary Coolant Average ROH Pressure kPa Reactor Thermal Pwr Reactor Generator Output Neutron Pwr System w Reactor Trip Turbine Trip 107 5 17 Loss of RC pump 1 This malfunction event causes one primary heat transport PHT pump to trip off line due to pump failure such as rotor failure The loss of one HTS pump will immediately initiate reactor power stepback gt Go to reactor coolant system screen insert the malfunction for loss of one PHT Pump P1 Observe that PHT Pump 1 is tripped off and the coolant discharge flow for P1 is decreasing rapidly Observe coolant flow in the other PHT pumps Observe that reactor power is stepped back Record the reactor power after malfunction is initiated Observe the coolant pressure and temperature transients Observe the turbine steam flow and turbine power Repeat the malfunction event again with the use of the reactor coolant system screen but before doing s
80. he core These are supplemented by fission chamber and ion chamber assemblies mounted in housings on the calandria shell The signals from the in core flux detectors are used to adjust the absorber insertion in the zone control assemblies By varying the absorber position in these assemblies the local neutron absorption in each zone of the reactor changes thereby controlling the local neutron flux level In contrast with CANDU 6 which uses liquid zones in core light water columns with varying liquid level as the zone control assemblies ACR uses control rods as the zone control assemblies Control absorber elements penetrate the core vertically These are normally parked out of the reactor core and are inserted to control the neutron flux level at times when a greater rate or amount of reactivity control is required than can be provided by the zone control assemblies Slow or long term reactivity variations are controlled by the addition of a neutron absorbing liquid to the moderator Control is achieved by varying the concentration of this neutron absorbent material in the moderator For example the liquid neutron absorbent material is used to compensate for the excess reactivity that exists with a full core of fresh fuel at first startup of the reactor In this regard two moderator poison addition systems are provided 10 a Boron addition initiated manually is used as a source of long term negative reactivity when the reactor ha
81. he disturbances caused by the turbine trip the plant control system is designed with the following control actions The reactor neutron power will be reduced quickly to 75 by rapid insertion of control rods this is known as reactor power setback The intent is to reduce the reactor power in ramp fashion but still maintain the reactor power at high enough level such that Xenon level buildup as a result of the setback will not overcome the positive reactivity margin available at the reactor power control system In other words at such reduced power level the reactor power control system still has enough positive reactivity from the rods to bring the reactor back to full power if the turbine trip can be cleared quickly The turbine bypass valves will open automatically when turbine trip is detected first the CSDVs then the ASDVs if SG pressure is still high trying to alleviate steam pressure build up After the reactor power setback has been completed the turbine bypass valves will modulate their opening to pass sufficient steam flow to the condenser in order to maintain SG pressure at the constant setpoint In this way the turbine bypass valves temporarily replace the turbine as the steam load and hence eliminate the mismatch of reactor thermal power and turbine power as mentioned previously To observe the transients as described above using the simulator gt First initialize the simulator to 10
82. hen the message will say MAN O P OK 58 The screen also provides a section for COOLANT INVENTORY FEED amp BLEED VALVES AUTO MANUAL CONTROL and BIAS Using pop up controls in this section one can switch the control mode AUTO MANUAL for DIRECT FEED VALVE BLEED VALVE CNA and CV5 The feed and bleed valves are normally in AUTO mode but may be placed on MANUAL and the valve opening can be controlled manually via pop up menus NOTE in order to control these valves MANUALLY one must use the pop up menu to switch the control mode from AUTO to MANUAL first then the control signal to the control valve will be frozen as shown in the numeric value display Observe the display message above the valve control If it says MAN O P OK that means the control valve can now be controlled by the MAN pop up menu If it says MAN O P NOT that means the MANUAL control signal from the MAN pop up and the frozen control signal to the control valve do not match One must then use the pop up menu to enter a value equal to the frozen numeric value display then the message will say MAN O P OK The amount of coolant feed and bleed is controlled about a bias value that is set to provide a steady flow of bleed to the purification system The amount of flow may be adjusted by changing the value of the BIAS by using the BIAS control pop up provided The last section on this screen makes provisions fo
83. hutdown systems These systems are designed to be both functionally different and geometrically separate The functional difference is achieved by the use of 24 shutoff units for SDS1 and six liquid injection nozzles for SDS2 The 24 shutoff rods are inserted vertically by gravity drop Their locations are shown in the Figure in section 2 2 The six poison injection nozzles are positioned horizontally as shown in same Figure indicated on the figure as through 6 A concentrated solution of gadolinium in 2 is injected under pressure into the moderator space between the calandria tubes The in core instrumentation feeding flux signals to the shutdown systems is also separated in a geometrical sense Vertical flux detector units and fission chambers on side D are used for SDS1 while horizontal flux detector units and ion chamber units on side B are used for SDS2 Other instrumentation monitoring the core conditions also feed into SDS1 and SDS2 Note SDS2 is not modeled in this simulator The display parameters shown on this screen are as follows The positions of each of the two SDS SHUTDOWN ROD banks are shown relative to their normal fully withdrawn position In this ACR Simulator the reactivity worth for each SDS SHUTDOWN ROD bank is 30 mk so the total reactivity worth for the two SDS SHUTDOWN ROD banks when fully inserted in core is 60 mk The trip time is 150 ms delay plus 1 5 sec for full SDS1 rods drop e REACTOR TR
84. ice pitch LP are shown below Core Configuration Total Fission Power Core Thermal Power Fuelling Scheme Radial Form Factor Peak Channel Power Peak Bundle Power Core average Exit Irradiation Core average Discharge Bundle Burnup Channel Average Exit Burnu Maximum Fuel Element Burnup Core average Dwell Time 0 95 NU CANDU Lattice LP 286 mm PTor 56 mm CTor 66 mm Vu Vr 164 ACR 700 Lattice LP 220 mm 58 mm CTor 78 mm Va Vi 7 1 PTor e ab st st ab ab ab ep a 7 P Oh o ge 2 2 Reactivity Control Units o E 2 CG amp e 2003 Zug 2019 am pu p 54 2 sus SU 14 CAU SU 19 e 9 903 2002 sue zous 501 2026 5016 ome EEN CAU 1 CAUS zeen 50 su SU 12 5017 UI 4 2001 a Z c c 501 pm suit SU 16 SU St Unis Fuel Sunde Positions emm 20 Zine Cem Uris Oink ACR 700 Reactor Core emm CAU Cortal Absorber Unis Vertical Detector Unis 284 Fuel Channels 220 mm Lattice Pitch Reactivity control units include neutron flux measuring devices reactivity control devices and safety shutdown systems Flux detectors are provided in and around the core to measure neutron flux and reactivity control devices are located in the core to control the nuclear reaction In core flux detectors are used to measure the neutron flux in different zones of t
85. ing system can be Windows 2000 or Windows XP The requirement of having a single PC to execute the models and display the main plant parameters in real time on a high resolution monitor implies that the models has to be as simple as possible while having realistic dynamic response The emphasis in developing the simulation models was on giving the desired level of realism to the user This means being able to display all plant parameters that are critical to operating the unit including the ones that characterize the main process control and protective systems The current configuration of the Simulator is able to respond to the operating conditions normally encountered in power plant operations as well as to many malfunctions as summarized in Table 1 The simulation uses a modular modeling approach basic models for each type of device and process to be represented as algorithms and are developed in FORTRAN These basic models are a combination of first order differential equations logical and algebraic relations The appropriate parameters and input output relationships are assigned to each model as demanded by a particular system application The interaction between the user and the simulator is via a combination of monitor displays mouse and keyboard Parameter monitoring and operator controls implemented via the plant display system at the generating station are represented in a virtually identical manner on the simulator Control panel i
86. ion to arrive at calibrated estimates of bulk and zonal reactor power The demand power routine DPR computes the desired reactor power setpoint and compares it with the measured bulk power to generate a bulk power error signal that is used to operate the reactivity devices The primary reactivity control devices are the 18 zone control absorber ZCU elements configured as nine units each containing two absorber elements The zone control absorber element insertions are varied in unison for bulk power control or differentially for tilt control In the Turbine Leads mode of operation the reactor power setpoint is calculated by the steam generator pressure control program SGPC In the Reactor Leads mode of operation the reactor power setpoint is set by the operator or in the case of abnormal plant conditions requiring power reductions is automatically calculated by the RRS program In addition to controlling reactor power to a specified setpoint the reactor regulating system monitors a number of important plant variables and reduces the reactor power when any of these variables exceed specified limits This power reduction may be fast stepback or slow setback depending on the possible consequences of the variable lying outside its normal operating range The signal processing logic associated with 3l RRS implemented in the distributed control system DCS is redundant and fail safe in software and hardware A genera
87. ions will appear Select the specific malfunction to initiate by clicking on the malfunction item itself The malfunction item will be highlighted in black gt Click on insert MF button if the malfunction is initiated immediately or input a time delay sec in the display box and then click insert the malfunction will be initiated after the specified time delay has elapsed When malfunction occurs the malfunction active alarm will be on clear a malfunction which has been inserted click on the malfunction item and then click clear MF or alternatively click on global clear which will clear all the malfunctions selected 5 1 Fail closed all feedwater level control valves This malfunction leads to loss of feedwater to the steam generators When this malfunction transient occurs The SG level drops quickly causing low SG level Reactor will be setback when SG level drops lt 10 11m Reactor will be tripped when SG level drops lt 9 9m Due to loss of feedwater to the steam generators cooling of the primary reactor coolant is reduced The higher temperature in the reactor coolant causes it to expand However as the reactor is tripped there will be rapid reduction of reactor thermal power causing shrinkage of reactor coolant So the net effect is the dropping of reactor coolant pressure Dropping coolant pressure causes out surge of coolant from the pres
88. ity Steam pressure MPa g Heat Transport Pumps Number Pump type Motor type Rated flow L s Rated head m Motor rating MWe ontainment Type Inside diameter m Height Top of base slab to Inside of Dome m Designed for LOCA Pressure kPa g MSLB Pressure kPa Turbine Generator Steam Turbine Type Steam Turbine Composition Net heat to turbine MWth Gross Net electrical output nominal MWe Gross Turbine Generator Efficiency Steam temperature at main stop valve Steam pressure at main stop valve MPa 2 Final feedwater temperature C Condenser Vacuum kPa a Vertical centrifugal single suction double discharge AC vertical squirrel cage induction Pre stressed concrete with epoxy liner 41 5 519 Hitachi impulse type tandem compound double exhaust flow reheat condensing turbine with a last stage blade length of 132 cm 52 inches One double flow high pressure cylinder two external moisture separators reheaters and two double flow low pressure cylinders 2062 CANDU 6 data quoted is based on the Qinshan Phase III CANDU 6 design 50 Hz Vertical U tube with integral preheater Vertical centrifugal single suction double discharge vertical squirrel cage induction Impulse type tandem compound double exhaust flow reheat condensing turbine with a last stage blade length of 132 52 inches One single flow high pressure cylinder two external moisture
89. l Pwr RCTR 105 0 PRIMARY COOLANT PRESSURE CONTROL TRML 80 0 waworPmorok PRESSURIZER HEATERS CONTROL 1 oo oe 5 oF 2 a ov 4 a or e mmo a oF Retr Outlet Pressure amp Setpoint p i PRESSURIZER PRESSURE RELIEF VALVES CONTROL UR PRZR Level amp Setpoint PRZR Relief Valves Pos cv22 0 Cv23 PRESSURIZER SPRAY VALVES CONTROL son aro res 0000 wor C EE scree oz zi C Eoee REACTOR COOLANT PRESSURE SETPOINT amp MODE CONTROL EL eee EL ILL ROH PRESSURE CONTROL MODE E NORMAL Coolant Pressure Resolution Time Scroll Reactor Outlet 12068 KPA ROH PRESS SETPOINT E 22100 kea CH f Max Max In Coolant Pressure Reactor Reactor Generator Output Primary Coolant Core Main STM Press Ferate Neutron Pwr Thermal Pwr Average ROH Pressure Flow kg s BOP STM Flow Contro Se WEE J FWFow EE EES S This screen is designed for reactor coolant pressure control The first section on screen provides controls for the pressurizer heaters The six HEATERS are normally in AUTO with the variable Heater 1 modulating The other five heaters are either ON or OFF and under AUTO control Via the pop up menus MANUAL operation can be selected
90. l block diagram of the reactor regulating system is shown in figure below 3 Mara Polion Addiion and Removal Reactor Parer Control Setback Powar Setpoint E enw Sisar Ganereler Praenum Danalar nut roe nen ame Reactor Regulating System Block Diagram Moderator Poison Control The reactivity depth of the zone control units ZCU in conjunction with the eight control absorbers MCA is sufficient to shut down the reactor even in the fresh fuel state when the fuel temperature reactivity feedback is at its maximum The normal method of maintaining the reactor adequately subcritical is by the manual addition of poison to the bulk moderator Addition of moderator poison is also a possible but unlikely means for the operator to reduce reactor power to low levels The poison solution is pre mixed in the respective tank and can be added under gravity to the moderator circulating pump suction line by opening a single valve from the poison tank via controls in the main control room The moderator poison system has a very large reactivity depth and is capable of reducing reactor power and keeping it adequately subcritical under any conditions Plant operating procedures define the level of moderator poison required to achieve the guaranteed shutdown state GSS in various circumstances Two moderator poison addition systems are provided 32 4 a Boron addition initiated manually is used as a s
91. l system to reduce or increase reactor neutron power in order that steam generator pressure will return to its setpoint Likewise under reactor leading control mode control signals will be sent to the turbine governor control system to reduce or raise turbine load in order that steam generator pressure will return to its setpoint In the event of a sudden turbine load reduction such as abnormal load rejection or turbine trip where the above described control system is not fast enough to alleviate steam pressure changes due to such transients an automatic steam bypass dump system is provided to dump the steam to the condenser and or to atmosphere if the steam generator pressure exceeds a predetermined setpoint Steam generator level control The steam generator level control system maintains a programmed water level that is a function of turbine load The control is a three element controller that regulates the feedwater valve by matching feedwater flow 1 element to steam flow 27 element from the steam generator while maintaining the generator level 3 element to its setpoint Turbine Load Control In the Turbine Leads mode the turbine load control can be done by the operator entering the target load known as Mega Watts MW Demand Setpoint and loading unloading rate This communicates its actions to the turbine generator governor controller through the steam generator pressure control SGPC program The turbine ge
92. low EE FwFow EE _ 115 Reactor Trip Turbine Trip 5 20 Primary Coolant 1 LOCA break This malfunction event causes a crack opening at the reactor inlet header 1 This break causes a loss of coolant accident LOCA event Before the malfunction is inserted it is recommended that the simulator user should be familiar with the design of the passive core injection system as described in Section 3 17 ACR passive core cooling screen before performing this exercise gt gt First load the full power initial condition IC and run the simulator Go to reactor coolant system screen and select the malfunction Primary Coolant RIH 1 LOCA break then press insert MF and press return Observe that the malfunction active alarm is on Note that all the trended parameters on the screen will change immediately Record the break flow in Table V Record the primary coolant pressure when the reactor is tripped After the reactor is tripped go to ACR passive core cooling screen On this screen the injection flow path by the passive core cooling system will be shown in thick blue lines during the various stages of injection cooling Record the parameters in Table VI during the various stages of injection Explain the coolant pressure transients in the course of event evolution When do press
93. main steam pressure setpoint The latter point requires the following explanation ACR 700 NPP is operating with a constant main steam pressure of 6 400 KPa Any mismatch of energy flow from the SG to the turbine will result in changes in the main steam pressure For example suppose at full power the turbine control valve is adjusted to 90 opening from its normal 100 opening The steam generators are generating 100 full power steam flow but the turbine only allows 90 steam flow to pass As a result the extra steam flow capacity in the SGs will increase the main steam pressure to a value higher than the current setpoint of 6 400 KPa The control system sensing that current main steam pressure is higher than the setpoint will signal the reactor regulating system RRS in lowering its reactor power setpoint accordingly As a consequence the primary coolant heat transfer to the SGs is reduced in a manner that would allow the SG main steam pressure to return to its setpoint 30 This mode of control is typically used for baseload operation with constant or scheduled load as well as load following operation with a frequency control function It is important to note that steam generator pressure is maintained constant during this control mode operation In the reactor leading control mode the reactor power control is determined by operator input and or plant upset conditions e g turbine trip which in turn will set a new reactor powe
94. mplemented by auxiliary systems which support its operation and maintain parameters within operation ranges to suit the various system functions The 3D isometric view of the Heat Transport System is shown below y e m i The pressure and volume of the coolant in the HTS are controlled by the pressure and inventory control system The long term cooling LTC system is used to remove decay heat following a reactor shutdown and to cool the HTS to a temperature suitable for maintenance of the heat transport and auxiliary system components The HTS and its auxiliary systems are similar to the equivalent systems in the CANDU 6 design However the overall design of these systems has been improved based on operation feedback from existing CANDU plants and has been simplified with the use of light water in a single loop configuration The major components of the heat transport system are the reactor fuel channels two steam generators four electrically driven heat transport pumps two reactor inlet headers two reactor outlet headers and the interconnecting piping Light water coolant is fed to the fuel channels from the inlet headers at each end of the reactor and is returned to the outlet headers at the opposite end off the reactor The principal function of the heat transport system main circuit is to provide reliable cooling of the reactor fuel under all operating conditions for the life of the plant and with minimal maintenance Th
95. n Maximum Flux Tilt error ising power cages Ier ed AK during power changes ipese RP mne e mee pe eener reegt qms ewe pc Gumww mw sorte ES Lows EES Neeser pep NOTE it may be necessary to record these values from the relevant trend in reactor power control screen or in the TRENDS screen 79 4 4 Turbine trip and recovery Turbine trip transient occurs as a result of either a load rejection or turbine malfunction On turbine trip The turbine main steam stop valves and governor valves will close immediately shutting off steam flow to the turbine As well the generator breaker will trip open causing the nominal MW power produced by the generator to drop to 0 MW almost instantly gt As a result of losing MW from the generator there is a large mismatch between the reactor thermal power and the turbine power at the SG This mismatch will cause a rapid increase in steam generator pressure which will cause disturbances to the reactor coolant system gt If action is not taken to reduce the reactor neutron power immediately the SG pressure safety relief valve will open on high SG pressure causing depressurization of the steam generator This again will cause disturbances in the primary systems To cope with t
96. n feedwater system feeds the steam generators from the condenser hotwell throughout the event The HTS flow decreases faster in the core pass downstream of the break If the break is large enough the flow will reverse in that pass For some break sizes the flow momentarily falls very low as the break upstream of the core pass balances the pumps Some channels may become steam filled and others may experience stratified two phase flow exposing some fuel elements to steam cooling Fuel temperatures rise A rise in fuel temperatures increases the internal fuel element gas pressures whereas a rise in sheath temperatures reduces the sheath strength Increased internal fuel element gas pressure along with the decreased coolant pressure increases fuel sheath stresses If the fuel sheath temperature becomes high enough sheath failure can occur e The pressurizer discharges its inventory into the HTS The decreasing pressurizer level causes the light water bleed valves to close and feed valves to open up adding light water makeup to the HTS e Following reactor trip the average fuel temperature decreases as the heat generation rate decreases and the temperature profile in the fuel pin flattens out The sheath temperature increases depending on the heat transfer from the sheath to the coolant e When the HTS pressure falls below a specified setpoint the ECI signal which is conditioned by the high reactor building pressure signal is generated This signal
97. ndenser level m bleed condenser pressure KPa Pressurizer level m and setpoint m Pressurizer spray flow kg s gt Coolant bleed flow kg s coolant feed flow kg s e Also shown on the screen is a pop up window for the Flow Diagram of the Feed amp Bleed circuit as shown below 56 57 3 12 ACR coolant inventory control Reactor Pwr amp Thermal Pwr ROH Pressure amp Setpoint 12500 105 0 RCTR PRIMARY COOLANT INVENTORY CONTROL TRML _PRIMMARY COOLANT INVENTORY CONTROL 80 0 10 01 57 PM 10 02 02 PM vu g Coolant Makeup Feed amp Bleed Valve Pos COOLANT INVENTORY FEED BLEED VALVES 1 AUTO MAN CONTROLS amp BIAS ep 100 0 SS FED Goerend ta Pos ORG Di P Ee Bleed Viv 5 Auro m Pos 916 man op manopnorok m Bleed viv cve 05 Sie j BLEED BIAS 500 40 0 20 0 REACTOR COOLANT PRESSURE CONTROL SETPOINT amp MODE ROH PRESSURE CONTROL MODE D NORMAL 12068 KPA ROHPRESSSETPOINT 12100 KPA 10 01 57 PM 10 02 02 PM Resolution Time Scroll 9 Max Out Max In Coolant Pressure Reactor Outlet oolant Inventory Reactor Reactor Generator Output Primary Coolant Core Main STM Press 6399 6 F 1 Neutron Pwr Thermal Pwr Average ROH Pressure Flow kg s reeze terate Control E SS BOP STM Fiow
98. nerator governor controller will regulate the steam flow through the turbine to meet turbine load target by controlling the opening of the turbine governor valve Emergency Core Cooling and Containment Control See details in section 2 7 c and d Reactor Shutdown Systems Control See details in section 2 7 a and b 35 The control loops described above can be summarized by the following control block diagram GENERATOR DEMANDED REACTOR POWER ELECTRICAL OUTPUT SETPOINT LEADS MODE ONLY TURBINE LEADS MODE 36 3 7 ACR Control Rods and Shutdown rods amp Reactivity 4 Average 8 ZCU s position REACTOR B 2 for flux tilt control aa TRIP KA Average 10 ZCU s position H for bulk flux control 49 13 wo Se ml GA E H D 250 a direction of coolant flow d SD RODS RESET 21 GR x A ve IN ve OUT 83 Avg ZCU Speed s 0 0001 MCA 1 Speed s 0 0000 Si A 2 Speed s 0 0000 d D RIH 1 Inflow ROH 2 Outflow A za 731 a 3567 0 3572 7 D QUE OUT a Ka D FUELT 714 8 2MCA 1MCA He 2M werage T Average RIH T 274 5 RIH 2 Inflow 3572 7 Average 326 7 1 1 6 0 5 0 40 A Is Reactor Reactor Generator Output Primary Coolant Core Main STM Press E Neutron Pwr 96 Thermal Pwr A
99. ng defined the Power Error and Flux Tilt the control algorithm for controlling the zone control units ZCU can now be described A digital control algorithm commonly known as velocity control algorithm is used to compute the speed of the respective zone control units ZCU according to its assigned control function namely for bulk flux control or for flux tilt control As described above the bulk flux control is mainly carried out by the 10 zone control rods located near the center of the reactor vessel namely Z2U Z2L 740 ZAL Z5U 751 760 Z6L Z U Z8L The spatial control is mainly carried out by the 8 zone control units ZCU located near the four corners of the reactor vessel namely ZIU ZIL Z3U Z3L ZTIU Z7L 290 Z9L The speed of a ZCUj in the 10 zone bulk flux control group is determined by evaluating the individual control functions as given below ZCU speed Reactor Bulk Power Control Resetting Control for the 8 ZCUs responsible for flux tilt control to 50 position Equalizing Control for the 10 ZCUS responsible for bulk flux control to achieve uniform position in core gt The speed of a ZCU in the 8 zone flux tilt control group is determined by evaluating the individual control functions as given below ZCU speed Reactor Bulk Power Control Flux Tilt Control Equalizing Control for the 8 ZCUs responsible for flux tilt control to achieve uniform position in core In co
100. nstruments and control devices such as push buttons and hand switches are shown as stylized pictures and are operated via special pop up menus and dialog boxes in response to user inputs This manual assumes that the user is familiar with the main characteristics of water cooled thermal reactor nuclear power plants as well as understanding the unique features of the CANDU 23 TABLE I SUMMARY OF SIMULATOR FEATURES SIMULATION SCOPE DISPLAY OPERATOR MALFUNCTIONS CONTROLS e Neutron flux levels ACR Reactor e reactor reactor setback and a range of 0 001 to Power control setpoint and rate stepback fail 110 full power 6 eACR Control of change inputleone bank of MCA delayed neutron groups Rods amp SDj to control rods drop into the Decay heat 3 groups rods computer reactor core e All reactivity controlje ACR Trip e manual control ofle al MCA rods devices zone control parameters reactivity devices stuck to manual rods ZCU absorber control rods rods MCA ZCU absorber gadolinium control rods MCA and e Xenon Iodine poison gadolinium e Spatial kinetic simulated addition removal for 18 reactor zones e reactor trip enabling display of flux e reactor setback tilt e reactor stepback e Reactor Regulating System RRS e Reactor Shutdown System SDSI e Main circuit coolant e Reactor e coolant heat epressurizer pressure loop with four pumps Coolant transport syst
101. nt and that the pressurizer spray comes in Observe the pressurizer level and record the bleed flow kg s Observe the bleed condenser level BLEED CDSR Level 5 17 29 PM il 5 17 29 PM PRZR Level amp Setpoint PRZR Spray Flow BLEED CDSR Pressure Feed Bleed i a Lee D ES MAKEUP 16 02 PM 106 Resolution Time Scroll FLOW DIAGRAM ROH PRESSURE CONTROL MODE 4l D NORMAL Dez CA NOTE Screen only shows typical arrangements Max Out Max In Not all valves are shown final design may change Core Main STM Press Pov een EE NE rur eg EES EE Primary Coolant lant Inventory amp Pressurizerw Reactor Trip Turbine Trip Reactor Reactor Generator Output Neutron Pwr Thermal Pwr 98 HTS process parameters at Pump Discharge ROH 1 1 m 4 Max Out E Max In Reactor Reactor Generator Output Primary Coolant Core Main STM Press WER Neutron Pwr Thermal Pwr Average ROH Pressure Flow kg s BOP STM Flow EEH KPa 99 5 12 RC inventory bleed valve CNS fails open This malfunction causes the reactor coolant bleed flow to reach the maximum As a result the bleed condenser level will increase The immediate impact on the reactor coolant system is decreased coolant inventory in the system The pressurizer level will decrease leading to decrease in pressurizer pressur
102. ntrol equation format the speed of the respective ZCU is given by For the 8 Zones ZCU _ SPD G1 Perr G2 FluxTC KL ZCUP AVG8ZCUP For the 10 Zones ZCU _ SPD G3 Perr GA Avg8ZCUP 0 5 G5 Avg10ZCUP ZCUP 46 Where ZCU SPD ZCU speed in 5 1 1 8 in the 8 zones group Gl gain constant for power error G2 gain constant for flux tilt error KL gain constant for the position equalizer for the 8 zones group ZCUP ZCU in the 8 zones group position 96 AVGS8ZCUP average ZCU position in the 8 zones group ZCU SPD ZCU speed in s j 1 10 in the 10 zones group G3 gain constant for power error G4 gain constant for resetting control of the 8 zones group G5 gain constant for the position equalizer for the 10 zones group ZCUP ZCUj in the 10 zones group position 96 AVGIOZCUP average ZCU position in the 10 zones group The ZCU control functions responsible for bulk flux control and flux tilt control respectively are illustrated by arrows in the following diagram extracted from the simulator screen gt The auto manual mode changeable by user ZCU average speed and the average position of the ZCUs are displayed respectively for the 10 zones group responsible for bulk power control and the 8 zones group responsible for flux tilt control The auto manual mode changeable by user absorber rods speed and the average position are displayed on
103. o first insert another malfunction for reactor setback amp stepback both failed The purpose is to study how the system thermal margin is challenged without the initial reactor power stepback Observe the reactor power transient coolant pressure and temperature transients reactor neutron power reactor thermal power turbine power transients Describe and explain the difference in responses when compared with the previous malfunction transient Discuss the thermal margin challenge in these cases and how the safety and control systems can cope with these challenges 108 Step Back Redd uus E RIH 1 2 ROH 1 2 Temp P1 2 3 4 Flows to RIH s 400 350 2 250 12 37 28 PM 12 39 45 12 37 28 PM 12 39 45 HTS process parameters at Pump Discharge 259 Tj 276 0 ROH 1 ROH 2 Pressure Feed Flow fd EEN 90 ve 51222015 12473 RIH 1 RIH 2 Pressure Reactor Power RIH 1 14000 j 2 8 130007 12000 11000 10000 m PM 12 39 45 1237 28 PM 12 39 45 Resolution Time Scroll Reactor Neutron Pwr e Max Out Max In FLOW DIAGRAM E Thermal Pwr Average ROH Pressure Flow kg s gop STM Flow EIS FW Flo Ww em Reactor Fee Generator Output Primary Coolant Core Main STM Press MODE G SETBACK STEPBACK TRIP Edu
104. open This malfunction leads to maximum feedwater flow to SG 1 with the control valve LCV 1 failed wide open Because the feedwater flow is much more than the steam flow from SG 1 as a result the level at SG 1 is rising steadily leading to SG 1 high level When this malfunction transient occurs Goto feedwater amp extraction steam screen observe that LCV 1 is 100 open gt Go to reactor coolant system screen observe SG 1 feedwater flow and steam flow Note the mismatch in flow and observe the SG 1 level gt Observe if this transient has any impact to the reactor and primary coolant systems As the SG level very high alarm occurs turbine generator will be tripped When the turbine is tripped there will be a Reactor Setback to 75 The transient response will be similar to that described in Section 4 4 6 WE LossRCPmps BOILER LEVEL SP Si 5 D NOT OK TURB sc Lv se v 13 2054 3 47 36 PM SG Level m 45 19 PM Time Scroll Resolution 4l J FLOW DIAGRAM Max Out Max In NOTE Screen only illustrates main feedwarer amp extraction steam flow arrangements Not all valves are shown See Flow Diagram for details Reactor Neutron Pwr Generator Output Primary Coolant Core Average ROH Pressure Flow kg s Main STM Press BETTER BOP STM Flow ZA FWFl
105. or power control system to reduce reactor power at the desired rate during power maneuvering due to the loss of control for the MCA rods When this malfunction transient occurs to reactor power control screen set the mode to reactor lead Enter target reactor power 70 and rate 0 5 per sec Accept the inputs gt Go to ACR Control Rods amp SD Rods screen observe the movement of ZCU rods and the MCA rods as the reactor power is decreasing towards the target power 70 gt Monitor the Power Error note the peak positive power error during this event Discuss the behavior of the Power Error how it increases and then decreases As you may notice all the ZCUs will be inserted fully in core initially where is the source of negative reactivity to reduce the reactor power to the target 70 FP Discuss MODE G SETBACK STEPBACK POWER REACTOR LEAD Current SETPOINT 10530 ae i 11 26 28 AM 11 28 02 Actual amp Demanded SP 26 D 110 00 MIN 0 00 0 15 0 11 26 28 AM 11 28 02 11 26 28 AM 11 28 02 aale LOAD 3 0006 ppm Top Bottom 7 0 111 Resolution Time Scroll FLUX side side 40018 D 77998 sus e d i Front Back 0 162 Max Out Max In Reactor Power Reactor Reactor Generator Output Primary Coolant Core Main STM Press Neutron Pw
106. orber Units Eight control absorber units MCAs are provided for rapid controlled power reductions and 38 to compensate for the fuel temperature reactivity effect for shutdown under fresh fuel conditions For the simulator the eight control absorbers are modeled as two banks of absorber rods The control absorber elements are physically similar to the shutdown SORs Normally the control absorbers are positioned outside the core Their arrangement is shown in the Figure shown in section 2 2 Since the reactivity increase following a power reduction is significant and usually rapid the zone controllers alone are incapable of counteracting the increase in all cases In particular the reactivity increase is the highest following a hot shutdown when fuel temperature drops to coolant temperature and for fresh fuel In this case MCAs are used to compensate for the reactivity increase The control absorbers are normally inserted in banks of two absorber elements each but can also be inserted individually The percentage insertion depends on the degree of reactor power reduction The optimum speed of insertion is determined primarily from control considerations In summary the maximum rate of positive reactivity insertion due to any set of reactivity devices of the reactor regulating system ranges between 0 05 mk s for MCAs and 0 2 mk s for the ZCU c Shutdown Systems The ACR 700 reactor is equipped with two physically independent s
107. ottom of each of the steam generator into the heat transport pumps two for each steam generator to repeat the cycle The system components and parameters shown on the screen are Average fuel temperature average coolant temperature C average core flow kg s A T across the core coolant outlet temperature coolant inlet temperature e Heat transport pump s discharge flow kg s discharge pressure KPa discharge temperature C 53 e Heat transport pump pop up control which allows START STOP and RESET operations e Pressure kPa flow kg s and temperature C at the reactor outlet header ROH 1 82 e For each steam generator SG feedwater flow kg s feedwater level in drum m steam drum pressure KPa main steam flow from SG to main steam header kg s e The following time trends are displayed RIH 1 2 ROH 1 2 temperatures C HTS pumps P1 P2 P3 P4 discharge flows Kg s to RIHs ROH 1 2 pressures The coolant feed flow kg s the coolant bleed flow kg s RIH 1 2 pressures KPa gt Reactor power Also shown on the screen is a pop up window for the Flow Diagram of the Heat Transport System as shown below STEAM GENERATOR 1 STEAM GENERATOR 2 HEAT TRANSPORT HEAT TRANSPORT PUMP PUMP LETHEADES P EAST REACIORGUILETHEADER TD UL REACTOR OUTLETHEADER 54 3 11 ACR coolant inventory and pre
108. ough bleed cooler then to the coolant purification system From it the coolant goes to the coolant makeup storage tank e Parameters displayed for the bleed condenser are pressure kPa temperature and level m The bleed condenser pressure is controlled via CV 14 by spraying cold coolant supplied from the feed pumps discharge line Furthermore the relief valve is available to relieve excessive high bleed condenser pressure e PRESSURIZER LEVEL SETPOINT and REACTOR OUTLET PRESSURE SETPOINT are also shown ROH PRESSURE CONTROL MODE control pop up is provided to facilitate the heat transport coolant pressure to be controlled in two modes NORMAL or SOLID SOLID mode represents the condition that the pressurizer is isolated from the heat transport circuit meaning that the isolating valve MV1 will be fully closed Therefore in SOLID mode there will be much pronounced pressure effects increase or decrease with changes in coolant mass inventory This mode is usually used during plant shutdown or cold startup when a fast coolant pressure decrease or increase is required In NORMAL mode as usually the case in normal plant operation the isolating valve 1 is fully open thus allowing the pressurizer to assist in maintaining coolant pressure and mass inventory at setpoint e The following time trends are displayed Pressurizer pressure KPa reactor outlet pressure average of ROH 1 2 pressures KPa Bleed co
109. ource of long term negative reactivity when the reactor has excess fuel reactivity Note Boron addition and removal is not modeled in the simulator b Gadolinium addition normally initiated manually is used as a source of short term negative reactivity to compensate for a lack of xenon gadolinium burns out at a rate similar to xenon production rate Under special conditions positive flux rate and large power error RRS will add gadolinium automatically Removal of poison from the moderator is via the moderator ion exchange purification facility Procedural controls ensure that no poison removal takes place during a deliberate shutdown state using moderator poison Successful poison addition requires the moderator circulating pumps to be operating Primary Coolant Pressure Control Reactor coolant pressure control in the ACR is performed by the pressurizer pressure control system This provides the capability of maintaining or restoring pressure at the design value following normal operational transients that would cause pressure changes It is done by the control of heaters and a spray in the pressurizer The system also provides steam relief capability by controlling the power relief valves Under normal operating conditions the pressurizer is the principal component in the pressure control of the HTS It is a pressure vessel that is partly full of liquid water with the remainder being saturated vapour in equilibrium with the liquid The P
110. ow E ED TY CT TTT WEEN Reactor Thermal Pwr 89 5 4 FW LCV 1 fails closed This malfunction leads to loss of feedwater to SG 1 As such the transient response is similar to that described in Section 5 2 5 5 Main Feed Water Pump trips This malfunction leads to loss of 50 of normal feedwater flow to SG 1 and SG 2 due to tripping of one SG feed pump BFP The result is low SG level causing reactor setback followed by reactor trip The transient response is similar to that described in Section 5 1 5 6 Turbine throttle PT fails low This malfunction causes the turbine throttle pressure transmitter to fail low The consequence is that the turbine governor control system is fooled into thinking that the main steam pressure is rapidly decreasing hence as a regulation control action the turbine governor will run back turbine load immediately in order to maintain main steam pressure Because the throttle pressure transmitter has failed low the turbine will be run back to 0 MW Turbine trip will follow as a consequence of generator zero forward power But in reality the main steam pressure was never low in the beginning Running back the turbine will cause immediate rise in main steam pressure Despite the fact that the turbine Bypass valves CSDV ASDV are opening to cope with the pressure rise it takes time for the steam pressure to decrease The peak rise in ste
111. r Thermal Pwr Pressure Flow kg s BOP STM Flow EZA euro WEE EE REECH ETH 1 105 5 16 Reactor setback stepback both fail This malfunction event impairs the first line of protective action initiated by the reactor control system RRS to decrease reactor power in response to process conditions that exceed the alarm limits However the reactor shutdown system SDS is always poised to act should those alarm limits reach the trip setpoint gt Go to control rods amp SD rods screen insert the malfunction reactor setback stepback both fail Use the pop up at bottom left to trip turbine Observe that due to the malfunction the reactor setback cannot be initiated therefore ZCU and MCA Control Rods will not respond to turbine trip Record reactor power after turbine trip Go to turbine generator screen observe the main steam pressure transient The turbine bypass valve CSDV ASDV should open to relieve steam pressure Go to reactor coolant system screen observe the transient in coolant pressure and temperature With the reactor setback stepback both failed is the safety margin e g coolant overpressure fuel temperature DNB etc of the system being challenged on a major transient like a turbine trip on loss of first line of reactor protective action malfunction of both the reactor setback and stepback Discuss the importance of s
112. r The extraction steam flows Kg s are shown respectively for turbine extraction as well as for main steam extraction to the deaerator e Main feedwater pump and auxiliary feedwater pump status with associated pop up menus for ON OFF controls HP heater motorized valves MV2 and pop up menus for open and close controls for controlling extraction steam flow to the HP heaters e Feedwater flow rate Kg s at SG level control valve LCV1 amp LCV2 outlet and feedwater temperature C e Pop up controls for auto manual for SG level control valves LCV1 amp LCV2 64 e Pop up controls for changing SG level setpoint control from computer SP to manual SP or vice versa gt NOTE in order to change the SG setpoint control from computer SP to manual SP one must use the pop up menu to switch the control mode from COMPUTER SP to MANUAL SP first then the steam generator level SP value will be frozen as shown in the numeric value display Observe the display message next to SP control status If it says MAN SP that means the SG level SP can now be controlled by the MAN SP pop up menu If it says MAN SP NOT OK that means the MANUAL SP value from the MAN SP pop up and the frozen SP value as displayed do not match One must then use the MAN SP pop up menu to enter a value equal to the frozen numeric value display then the mess
113. r changing the reactor outlet pressure setpoint and the pressure control mode for the heat transport system The current reactor outlet pressure is shown and the reactor outlet pressure setpoint KPa may be controlled manually via the control pop up provided As well PRESSURE CONTROL MODE control pop up is provided to facilitate the heat transport coolant pressure to be controlled in two modes NORMAL or SOLID SOLID mode represents the condition that the pressurizer is isolated from the heat transport circuit meaning that the isolating valve MV1 will be fully closed Therefore in SOLID mode there will be much pronounced pressure effects increase or decrease with changes in coolant mass inventory This mode is usually used during plant shutdown or cold startup when a fast coolant pressure decrease or increase is required In NORMAL mode as usually the case in normal plant operation the isolating valve MV1 is fully open thus allowing the pressurizer to assist in maintaining coolant pressure and mass inventory at setpoint The following time trends are displayed Reactor neutron power 96 reactor thermal power 96 Reactor outlet header pressure average of 1 2 pressure KPa amp setpoint KPa gt Pressurizer level m amp setpoint m gt Reactor coolant makeup feed valve position reactor coolant bleed valve position 59 3 13 coolant pressure control Reactor Pwr amp Therma
114. r setpoint The water steam system consisting of the turbine with its bypass system and the steam generators will adjust turbine load MW and or other steam loads such as steam dump to atmosphere or condenser to match with any reactor power changes whilst maintaining the steam generator pressure constant In support of these two control modes and plant safety functions the ACR has the following control loops as illustrated by the ACR control loops screen in the simulator 1 2 Reactor power demand SP Reactor power demand setpoint SP is determined by operator input and or by the automatic limitation functions such as the reactor stepback which requires a step change in power reduction or reactor setback which requires power reduction at a fixed rate The automatic limitation functions are triggered by specific reactor coolant process conditions which exceed alarm setpoints The Reactor power demand setpoint SP provides input to the computer control program demand power routine DPR Reactor Regulating System RRS The reactor regulating system RRS is composed of input sensors fission chambers in core flux detectors and process measurements reactivity control devices zone control units control absorbers hardware interlocks and display devices The power measurement and calibration routine uses measurements from a variety of sensors self powered in core flux detectors fission chambers process instrumentat
115. r the absorber banks is as follows If the absorber banks control is set in AUTO the absorber banks will move according to the power error versus zone control rods position as per the above Reactivity Limit Control Diagram In the GREEN color region designating Reactor Power Error as PE amp Average Zone Control Rods Position as ZCP the green color region is defined by a 3 2 2 4 80 ZCP 0 and b 4 gt PE 2 5 100 ZCP 2 0 96 In this region the absorber bank 2 will be driven OUT first if it is in core and absorber bank 1 will start to drive OUT when bank 2 is completely driven out In the LIGHT BROWN color region it is defined by a 5 2 PE 2 396 100 gt ZCP gt 0 96 and b 3 2 PE 2 4 100 ZCP gt 85 In this region the absorber bank 1 will be driven IN first and bank 2 will start to drive IN when bank 1 is completely driven in core In the DARK BLUE region it is defined by 7 96 2 PE gt 5 100 2 ZCP 2 0 In this region both banks of absorber rods will be driven OUT simultaneously In the MAGENTA color region it is defined by 41 7 gt gt 5 100 2 ZCP 2 0 In this region both banks of absorber rods will be driven IN simultaneously e Inthe LIGHT BLUE region it is defined by 3 2 PE 2 4 85 ZCP 80 96 This is a transitional region between the GREEN region and the LIGHT BROWN region where the absorber rods which are
116. ressurizer is connected to the Reactor Outlet Header ROH of the HTS via a motorized valve When this motorized valve is fully open the control is in Normal mode Should any pressure changes occur in the reactor coolant outlet header in the HTS there will be in surge or out surge of coolant into or out of the Pressurizer through this motorized valve depending on the differential pressure between the Pressurizer and ROH Hence it is necessary to provide the capability of maintaining or restoring pressurizer at the design pressure value following normal operational transients which would cause Pressurizer pressure changes This Normal mode of pressure control is handled by the pressurizer pressure control system which controls a number of immersion heaters as well as a spray system in the pressurizer The system also provides steam relief capability by controlling the steam relief valves to the Bleed Condenser At low reactor powers the pressurizer may be isolated from the HTS by closing the motorized valve that is normally open to connect HTS to the pressurizer In this case it is known as the Solid mode of pressure control In solid mode the pressure of the HTS measured at the reactor outlet headers is controlled by adjusting the feed and bleed flows in and out of the HTS 33 5 Primary Coolant Inventory amp Makeup control The primary coolant inventory amp makeup control is performed by the pressurizer level control system It p
117. rno RAE 68 4 ACR BASIC OPERATIONS amp TRANSIENT RECOVERY eeeee esee eee 72 4 1 PLANT LOAD MANEUVERING REACTOR LEAD ccscssssssececeesessscecececeesenseseceeccecsessaaececececsersaeeeeeeeeeeneas 72 4 2 PLANT LOAD MANEUVERING TURBINE LRAD 74 4 3 POWER LEVEL REDUCTION TO 0 FP ou ccccssssccececsessssececececsenecaececececsesaaececececeeaaeeeeececsessaaeeeeeeeeeeneas 78 4 4 TURBINE TRIP AND RECOVERY cech 80 4 5 REACTOR TRIP AND eicere e eere eterne aeree ECO 82 5 ACR MALFUNCTION TRANSIENT EV ENTIS ecce ee ee sesto seta sese sns toss seen s sane sena 85 5 1 FAIL CLOSED ALL FEEDWATER LEVEL CONTROL VALVES tenant seen 85 5 2 STEAM GENERATOR 1 STEAM FLOW FT IRRATIONAL 88 o IFWECEVAET BATES OPEN aeri eee ed eee reete ete decavnestentded ee eire tte bd oe bee ted Ue rede 89 DAS IFW ECWMHEBAIES CEOSED tr EE RU ete ee Ee 90 S3 MAIN C 90 5 6 TURBINE THROTTLE PT FAILS LOW ccccsscccccceceessaececccecsessaececececsessaececececeeseasceceesceesensaceseeseeeneneaeees 90 5 7 ALL ATMOSPHERIC MSSVS FAIL OPEN 92 5 8 TURBINE BYPASS VALVE CSDV FAILS CLOSE 94 5 9 TURBINE SPURIOUS TRIER E Soave pire tree ee Pec
118. rovides the capability of establishing maintaining and restoring the pressurizer water level to the target value which is a function of the average coolant temperature affecting coolant swell and shrink It maintains the coolant level in the pressurizer within prescribed limits by adjusting the flow of the coolant feed and coolant bleed system thus controlling the reactor coolant water inventory Steam ERI Heat Ti rt Syst eat Transport System Inlet pm Heat Transport System Outlet Pa Steam Generator Heat Transport System e Pressurizer 27 Pump Assembly ump Assembly Fuel Channel 21 Heat Transport System Inlet Header 34 6 7 8 9 10 11 MW demand setpoint demand Megawatts MW demand setpoint is determined by operator input This input will used as reference target for raising or lowering the turbine load under turbine lead mode Steam Generator Pressure Control SGPC Steam generator pressure is maintained at an equilibrium constant value determined by the heat balance between the heat input to the steam generator and the turbine steam consumption If during power maneuvers or plant upset there is a mismatch between reactor thermal power and the turbine power steam generator pressure will vary and deviate from the pressure setpoint Under turbine leading control mode control signals will be sent to the reactor power contro
119. s etii tudin mts 16 3 700 MW E ADVANCED CANDU REACTOR NPP SIMULA TOR e esee 23 3 1 SIMUDATOR STARTUR S4 tete eine tt 26 3 2 SIMULATORSINITIALIZATION ste eet iN lev EH pe UO LE S 26 3 3 LIST OF ACR SIMULATOR DISPLAY SCREENS nenne 26 3 4 SIMULATOR DISPLAY COMMON PRATURES a dass se eterna rena annees 27 3 5 ACR PLANT OVERVIEW Eege ied EIU RT D Ie a 28 3 6 30 3 7 ACR CONTROL RODS AND SHUTDOWN RODS amp REACTIVITY isses ennemi 37 3 8 ACR REACTOR POWER CONTROL A E E aE 44 3 9 TRIP PARAMETERS eot tei nen washes 50 3 10 ACR REACTOR COOLANT SYSTEM eege tessile re Eve ve ertet Ee ba 53 3 11 ACR COOLANT INVENTORY AND PRESSURIZER en 55 3 12 ACR COOLANT INVENTORY CONTROL enne ses ette tensa sese nna 58 3 13 ACR COOLANT PRESSURE CONTROL csssesssseeececsessscececcceesenseaeceeececsessaaececececeeaaesecececsessaaeeeceeeeeeneas 60 3 14 ACR TURBINE GENERATOR ee eee eee eee pee e peo P perte ipee et 62 3 15 ACR FEEDWATER AND EXTRACTION STEANM emen enne thnn nsns enint rte assess eene ten ns 64 3 16 ACR MW DEMAND SETPOINT SP AND STEAM GENERATOR PRESSURE CONTROL SGPC 66 3 17 AGR PASSIVE CORE COOLING 55 breiter te
120. s excess fuel reactivity b Gadolinium addition normally initiated manually is used as a source of short term negative reactivity to compensate for a lack of xenon gadolinium burns out at a rate similar to xenon production rate Under special conditions positive flux rate and large power error Reactor Regulating System RRS will add gadolinium automatically Two independent reactor safety shutdown systems are provided Each shutdown system acting alone is designed to shut the reactor down and maintain it in a safe shutdown condition The safety shutdown systems are independent of the reactor regulating system and are also independent of each other The first shutdown system SDS1 consists of shutoff units absorber element guide assembly and drive mechanisms which drop neutron absorbing elements into the core by gravity on receipt of a shutdown signal from the safety system The second shutdown system SDS2 uses injection of a strong neutron absorbing solution into the moderator The automatic shutdown systems respond to both neutronic and process signals 2 3 Heat Transport System The heat transport system HTS circulates pressurized light water coolant through the reactor fuel channels to remove heat produced by nuclear fission in the core The fission heat is carried by the reactor coolant to the steam generators to produce steam on the secondary side that subsequently drives the turbine generator The heat transport system is co
121. separators reheaters and two double flow low pressure cylinders Gross electrical output is dependent on cooling water temperature the turbine generator and condenser design and the grid frequency 60 Hz 2 BRIEF ACR 700 SYSTEMS OVERVIEW 2 1 Reactor Configuration Calandria Vault Reactivity Mechanisms Deck Reactivity Unit Thimbks Calandria Relief Duct Fission Chambers Shield Cooling Inlet Header Mogcerator Inlet End Shield LISS Units Shielding Balls Calandria lon Chambers Fuel Channel End Fittings Calandria Support P Reactivity and Embedment p Absorber Guides The ACR reactor consists of a set of 292 horizontally aligned fuel channels arranged in a square pitch The fuel channels contain the fuel and the high pressure light water coolant They are mounted in a calandria vessel containing the heavy water moderator Individual calandria tubes surround each individual fuel channel The calandria vessel is enclosed by end shields which support each end of the calandria They are filled with shielding balls and water to provide shielding The fuel channels are located by adjustable restraints on the two endshields and are connected by individual feeder pipes to the Heat Transport System The calandria vessel is enclosed in a concrete vault calandria vault filled with light water for shielding The calandria vault is closed at the top by the reactivity mechanisms deck The core configuration parameters and latt
122. sociated with the turbine and the generator The parameters displayed are Main steam pressure KPa and main steam flow Kg s main steam stop valve MSV status Main steam header pressure KPa e Status of main steam safety relief valves MSSVs e Control status auto manual opening and flow Kg s through the steam bypass valves CONDENSER STEAM DISCHARGE VALVES CSDV ATMOSPHERIC STEAM DISCHARGE VALVES ASDV e Steam flow to the turbine kg sec e Governor control valve position CV open e Generator output MW station services MW e Turbine generator speed of rotation rpm e Generator breaker trip status e Turbine trip status tripped or reset e Turbine control status auto by computer or manual 62 The trend displays are gt Reactor neutron amp thermal power gt Generator output MW Turbine steam flow Kg s steam BYPASS flow Kg s Turbine speed RPM Turbine governor position gt Main steam stop valve MSV inlet pressure KPa The following pop up menus are provided TURBINE RUNBACK sets target and rate sec of runback when accept is selected TURBINE TRIP STATUS trip or reset Steam bypass valve AUTO MANUAL control AUTO select allows transfer to MANUAL control following which the manual position of the valve may be set Computer or manual control of the speeder gear Turbine runup speedup controls 63 3 15 fe
123. splay under the trends needed to withdraw all shutdown rods As well observe the position of the yellow cursor show on the Reactivity Limit Diagram after the SD rods are fully withdrawn and the RRS is taking full control You may observe that initially the RRS will input a small power increase setpoint of 1 45 FP see ACR Reactor Power Control Screen in order to pull out the ZCU and MCA rods Observe the movement of the yellow cursor and the corresponding movement of the ZCU rods and MCA rods As well go to ACR Reactor Power Control Screen observe the trends for overall mk average ZCU and MCA position Can you explain the trends for overall mk and the average ZCU and MCA position e Raise reactor power to 60 FP in small step at rate of 0 3 sec e Observe the response of the reactor regulating system and the reactivity changes that take place 82 Now call up the 100 FP IC then call up the IC ZCU DOUBLE WORTH meaning that the ZCU s now have a total MK worth of 18 MK instead of 9 MK Repeat the REACTOR TRIP amp RECOVERY exercise using this new MK worth of ZCUs Discuss and compare the results for these two exercises 83 84 5 ACR MALFUNCTION TRANSIENT EVENTS Note The ACR malfunction transient events described below are caused by malfunctions initiated in the simulator To initiate a malfunction Press the MALF button at the bottom right of any screen A pop up menu with a list of malfunct
124. ss 5765 0 Neutron Pwr Thermal Pwr Average ROH Pressure Flow kg s BOP STM Flow a FW Flow Emm ee ent aaa a ES Hi Neutron Pwr ROH Press Hi Hi Stm Gen Level Lo PRZR Lvl Hi Low Fwd Pur Trip Main BFP s Trip Hi Neut Pwr LogR ROH Press Hi Main Stm Pres Hi Stm Gen Level Hi PRZR RCTR OUT Pressure BLEED CDSR Level 12150 Gg 1 31 08 PM Coolant Feed amp Bleed Flows 1 31 08 PM 1 28 51 PM 1 31 08 PM Resolution Time Scroll FLOW DIAGRAM ROH PRESSURE CONTROL MODE A e NOTE Screen only shows typical arrangements Max Out Max In DI NoRMAL Not all valves are shown final design may change Reactor Laud Reactor Generator Output Primary Coolant Core Main STM Press Neutron Pwr Thermal Pwr Average ROH Pressure Flow kg s BOP STM Flow KPa SS Transients for the loss of PHT P1 and P3 HTS process parameters at Pump Discharge edel su ovs Mu 0 Ave cone flow ess Too zs wenn es Be Reactor Generator Output Primary Coolant Core Main STM Press System Neutron Pwr Thermal Pwr Average ROH Pressure Flow kg s BOP STM Flow v KPa RES FW Flow a Ten 113 5 19 100 main steam header break This malfunction event causes steam pipe break in the main steam line before the main steam stop val
125. ssurizer PRZR RCTR OUT Pressure BLEED CDSR Level ER PRZR Level amp Setpoint 3 04 46 PM 0434 3 04 46 Coolant Feed amp Bleed Flows BLEED CDSR Pressure 5000 4000 3000 2000 1000 3 04 46 PM B 04 34PM 3 04 46 PM 1 D MAKEUP x Resolution Time Scroll FLOW DIAGRAM ROH PRESSURE CONTROL MODE 4 D NORMAL NOTE Screen only shows typical arrangements Max Out Max In Not all valves are shown final design may change oolant Inventory Reactor Reactor Generator Output Primary Coolant Core Main STM Press amp a Neutron Pwr 96 Thermal Pwr Average ROH Pressure Flow kg s BOP STM Flow EE Freeze Tterate Pressurizer w KPa FWFiow RER eoo o REES RER RER EEN 7 SE This screen shows the coolant inventory and pressure control system including the pressurizer pressurizer pressure relief coolant feed and bleed circuits bleed condenser bleed cooler coolant purification system and coolant makeup storage tank e Starting with the coolant makeup storage tank at the bottom left hand corner its level is displayed in meters The tank supplies the flow and suction pressure for the feed pumps P1 and P2 normally one pump is running the pop up menu allows START STOP and RESET operations e The flow kg sec and temperature C of the coolant feed flow are displayed The feed flow then passes through
126. sure and temperature Record the reactivity feedback effects mk for Xenon What is the difference in mk for Xenon before amp after the turbine trip e Go to turbine generator screen reset turbine trip select TRU ENABLE and select TRU speedup to synchronize the generator and continue to load turbine e After turbine is in service what happens to the steam bypass valves as the turbine power increases Note the SG pressure reading e After the turbine power is equal to the reactor power go to the reactor power control screen to increase reactor power to 100 in 25 steps at 0 5 per sec Now call up the 100 FP IC then call up the IC ZCU DOUBLE WORTH meaning that the ZCU s now have a total MK worth of 18 MK instead of 9 MK Repeat the TURBINE TRIP amp RECOVERY exercise using this new MK worth of ZCUs Discuss and compare the results for these two exercises 81 4 5 Reactor trip and recovery Reactor trip or reactor scram is a reactor protective action initiated by the reactor safety shutdown system on detection of alarm limits exceeded by specific parameters in the reactor core coolant and balance of plant systems The parameters and the related reactor trip setpoints are described in Section 2 9 ACR trip parameters Most importantly the reactor also can be tripped by the operator MANUALLY on account of abnormal incidents or accidents The reactor trip action is to drop the two banks of
127. surizer in order to alleviate coolant pressure decrease Observe the flow direction in the surge line to pressurizer As well the electric heaters in the pressurizer will be turned on until coolant pressure returns to its setpoint 85 As reactor is tripped SG pressure is dropping rapidly causing the turbine governor to runback the turbine that is closing the turbine governor control valves This results in rapid reduction of MW to zero leading to turbine generator trip on zero forward power Press o Lo nnlant Flew a RCTR Neut Thm Pwr Tara Pot D A stm Hews Extracton Main E BOILER LEVEL SP SP cow se D sc wv en v 11 6469 D A Pressure amp SP Main Sun Hl Pressure DAP 602 2 8000 2 54 52 PM 52 55 9M 2 54 52 PM WS Ext Stm Flows SG Leval m 16 SG1 SG2 2 52 35 FM 2 54 52 FM p 52 35 2 54 52 9M SE i Tinz Sc ul I yos i FLOW DIAGRAM Max Out Scie dt duslieles meir amp esli stion seen uve Se arrangements Not al valves are showr ses How Dacram for dears Aeactor Reactor Generator Gutput Primary Cooant Core Man SiMPress Neutron Dwr 25 Thermal Pwr Average ROH Dress ms Flow kgs DOP 3 Flan ES EW Flow Fuel Temp RIH 1 2 ROH 1 2 Temp P1 2 3 4 Flows to RIH s 1 RIH 1 M 400 2 RIH 2 8 300 3 ROH 18 on 4 2 100 52 35 2 54
128. t of these very effective hands on educational tools It is important to remember however that the application of these PC based simulators is limited to providing general response characteristics of selected types of power reactor systems and that they are not intended to be used for plant specific purposes such as design safety evaluation licensing or operator training The IAEA simulator collection currently includes the following simulators e WWER 1000 simulator provided to the IAEA by the Moscow Engineering and Physics Institute in Russia e The IAEA generic Pressurized Water Reactor PWR simulator has been developed by Micro Simulation Technology of USA using the PCTRAN software This simulator is a 600 MWe generic two loop PWR with inverted U bend steam generators and dry containment system that could be a Westinghouse Framatome or KWU design e The IAEA advanced PWR simulator has been developed by Cassiopeia Technologies Inc CTI of Canada and is largely based on a 600 MWe PWR design with passive safety systems similar to the Westinghouse AP 600 e The IAEA generic Boiling Water Reactor BWR simulator has also been developed by CTI and represents a typical 1300 MWe BWR with internal recirculation pumps and fine motion control rod drives This simulator underwent a major enhancement effort in 2008 when a containment model based on the ABWR was added e The IAEA Pressurized Heavy Water Reactor PHWR simulator is also a CTI
129. team bypass system as well as the main steam safety features 106 eme ENENENNM o WE Main Trip j_Hinewtewriogt wesen anser sincenteveis Las L spe RCTR Neut Thrm Pwr Main Steam Header P Generator Output MW To HP heaters STATION 800 0 66947 SERVICES 5900 Mw T MSV Inlet dl 6567 6 GENERATOR OUTPUT gross _0 00 mw 11 50 45 AM 11 53 04 11 50 45 AM 11 53 04 Steam CSDV ASDV MSSV Flow Turbine Speed 1200 0 1900 11 53 04 11 50 45 AM 11 53 04 MSV Inlet Pressure MSSV1 TURBINE TRIP STATUS J MSSV2 des 60 0 flow Com TRIPPED Ll Speeder Gear 96 om 40 0 20 0 SPDR GEAR i 0 0 CONTROL AUTO D 11 50 45 11 53 04 11 50 45 TURBINE T Resolution Time Scroll 4 RUNBACK Max Out In NOTE Screen only illustrates main steam arrangements Core Flow kg s Not all valves are shown See Flow Diaaram for details Reactor Reactor Generator Output Primary Coolant Neutron Pwr Thermal Pwr Average ROH Pressure STM Flow ccs rosse BERE NE NN 7239 55 Matt Help 1 Fuel Temp FW Flow z reo BEE op ss H Neut Pwr LogR Stm Gen Level Loss RC Profs RIH 1 2 ROH 1 2 Temp P1 2 3 4 Flows to RIH s HTS process parameters at Pump Dischar
130. the feed isolation valve MV18 before entering Steam Generator 2 at the suction point of the coolant heat transport pumps P2 P4 Note that some of the coolant at the feed circuit is heated up through the bleed condenser reflux line via CV11 The heated feed coolant then mixes with the coolant from CV12 discharge before the mixture passes through MV 18 e Coolant flow from the reactor outlet header ROH 2 is normally to and from the pressurizer via a short connecting pipe a negative flow kg sec indicating flow out of the pressurizer into ROH 2 Vice versa would indicate a positive flow Pressurizer pressure kPa temperature and level m are displayed Pressurizer pressure is maintained by one variable and five on off heaters which turn ON if the pressure falls and by pressure relief valves CV22 and CV23 are open if the 55 pressure is too high As well coolant is drawn from connecting lines with the reactor inlet header RIH 2 via control valves SCV1 SCV2 for the purpose of spraying to depressurize the pressurizer There is coolant bleed flow kg sec from the steam generator SG 1 outlet cold coolant suction lines of heat transport pumps P1 P3 The coolant bleed flow via the bleed control valves CV5 CV6 and isolating MV8 will help maintain coolant inventory in the main coolant circuit if the inventory becomes too high as sensed by high pressurizer level e The outflow from the bleed condenser goes thr
131. the sumps into the HTS via the LTC heat exchangers The LTC delivers flow to the reactor inlet headers thereby utilizing the cooling path already established by the high pressure ECI system The LTC system is also used for long term cooling of the reactor after shutdown following other accidents and transients The followings provide a qualitative description of the ECC event sequence That is the event sequence describes the behavior that would be expected should a LOCA occur e A large break is postulated to occur in a large diameter pipe of the heat transport system HTS discharging coolant into containment 69 e The pressure temperature and humidity of the containment atmosphere increase The HTS depressurization causes coolant voiding in the core and a decrease in reactivity The reactor shuts down on a process trip e g low Heat Transport System pressure low Heat Transport System flow depending on break size and initial reactor power Containment isolation is automatically initiated on a high reactor building pressure signal The high reactor building pressure signal also conditions ECI signal The heat transport system loses inventory and depressurizes at a rate depending on the break size and location Following reactor trip the turbine runs back The condenser steam dump valves CSDVs open to by pass steam to the condenser The atmospheric steam discharge valves ASDVs open and close to maintain system pressure e The mai
132. then recovers 96 Turbine Trip ROH Press Lo Lo Step Back Req d Setback Redd Gen Breaker Opn ROH Press Hi Hi Coolant Fon Lo ROH Press Hi Loss 5 unction Active PRZR RCTR OUT Pressure BLEED CDSR Level 5 05 06 PM Coolant Feed amp Bleed Flows BLEED CDSR Pressure EE mu Resolution Time Scroll FLOW DIAGRAM ROH PRESSURE CONTROL MODE 4 D NORMAL D NOTE Screen only shows typical arrangements Max Out Max In Not all valves are shown final design may change lant Inventory Reactor Reactor Generator Output Primary Coolant Core Main STM Press Freeze Tterate amp P Neutron Pwr Thermal Pwr Average Flow kg s BOP STM Flow 4160 9 rwro RER 5 7 LY TT TET TE 7 Ma Hee ESCH LHe Casse eps _smcenteveirs Lesen Lespe mr RIH 1 2 ROH 1 2 Temp P1 2 3 4 Flows to RIH s HTS process parameters at Pump Discharge Tj 280 8 5 02 01 PM 5 04 18 PM 5 02 01 PM 5 04 18 PM ROH 1 ROH 2 Pressure es Feed Bleed Fow e He ven __ D 51 1 12500 0 50201 5 04 18 5 04 18 1 RIH 2 Pressure Reactor Power RIH 1 14000 RIH 2 13500 13000 25 12500 5 02 01 PM 5 04 18 PM 5 02 01 PM 5 04 18 PM Resolution Time S
133. ual Setpoint Demanded Power Setpoint Demanded Rate Setpoint Current Reactor Power Average Coolant Temperature from ACR Reactor Coolant Screen Average Core Flow Average Fuel Temp 72 Coolant Delta C ROH temp RIH temp Reactor Outlet KPa Header 1 Pressure Reactor Outlet KPa L i l Header 2 Pressure Reactor Inlet Header 1 Pressure Reactor Inlet Header 2 Pressure Pressurizer Level Flux Tilt Top bottom Side side Front back ZCU 8 zones group control rods average position ZCU 10 zones group control rods average position MCA Bank 1 rods average position MCA Bank 2 rods average position Gadolinium ppm Concentration Pressurizer Main Steam Pressure Total steam flow Kg s from steam generators Total Feedwater Kg s Flow 73 5 6 7 8 9 4 2 Reduce power using reactor power setpoint pop up gt Press the reactor power setpoint pop up button at the bottom left corner of the screen gt Enter reactor power SP target 90 enter power rate 0 3 sec and press accept Observe parameter changes during transient and record comments Freeze simulator as soon as reactor neutron power just reaches 90 and record parameter values in the column 2 for 90 power just reached Unfreeze simulator and let parameters stabilize record parameter values in the column 3
134. ual stepback which is done from this page on the simulator the target value 96 needs to be input gt TRIP status is indicated by YES or NO reactor trip is initiated by the shutdown system if the condition clears it can be reset from here Note however that the tripped shutdown system must also be reset before RRS will pull out the shutdown rods this must be done on the shutdown rods page 44 Key components of RRS control algorithm are also shown on this screen gt REACTOR POWER SETPOINT target and rate are specified by the user on the simulator in terms of FP and FP sec i e as linear measurements instead of the logarithmic values used in practice The requested rate of change should be no greater than 0 8 of full power per second in order to avoid a reactor LOG RATE trip This is readily achieved in the at power range above 15 FP but only very small rates should be used at low reactor power levels below 1 FP such as encountered after a reactor trip The MW DEMAND SETPOINT is set equal to the MW SETPOINT under TURBINE LEADING control the upper and lower limits on this setpoint can be specified here The ACTUAL SETPOINT is set equal to the accepted REACTOR POWER SETPOINT TARGET under RRS control in REACTOR LEADING mode HOLD POWER On will select REACTOR LEADING mode and stops any requested changes in DEMANDED POWER SETPOINT DEMANDED RATE SETPOINT is set equal to the accepted REA
135. uction of heavy water inventory 72 less D O mass inventory when compared with CANDU 6 6 30 Inside Diameter 480 Channels 3 Individual Fuel Channel 5 20 Inside Diameter 299 284 Channels DOG Annulus Gas CO3 Fuel Element Pressure Tube Fuel Sheath Coolant Calandria Tube The design also features higher pressures and temperatures of reactor coolant and main steam coolant outlet temperature 326 deg C and reactor inlet header RIH pressure 13 MPa thus providing an improved thermal efficiency than the existing CANDU plants In particular the use of the CANFLEX fuel bundle with lower linear rating and higher critical heat flux permits increased operating and safety margins of the reactor average channel power increased from 5 3 MW CANDU 6 to 6 8 MW ACR 700 The details of the CANFLEX fuel bundles in an ACR 700 core are illustrated below Reactor core configuration with 284 fuel channels 2CANFLEX fuel bundles per channel 43 fuel elements in one CANFLEX fuel bundle The bundle has two elements size centre pin and inner ring of seven elements with a diameter of 13 5 mm The outer two rings consist of 35 elements with 11 5 mm diameter 2 1 wt U 235 in 42 pins The center pin contains burnable poison U Dy O2 pellet with 7 596 wt Dysprosium in natural Uranium CANFLEXO is a registered trademark of AECL and the Korea Atomic Energy Research Institute KAERI e CANDU
136. ure bumps occur And why do they occur Explain why the accumulator is necessary Can the accumulator be eliminated if we make use of the Reserve Water Tank RWT located at the top of the Containment building Explain why the opening of the MSSV is necessary to serve what purpose 116 TABLE VI 1 BREAK Stages of Accumulators Steam Reserve Water Long Term Cooing LTC Injection n service Generators Tank in service in service Crash Cool MSSVs after Break Break after Break Break III LY Turbine power Reactor Thermal Power Kg s ET FP Flow Kg s Kg s Average Coolant Temp CC EelTepQO E PRZRlevel m III PRZR Pressure KPa Coolant Pressure at RIH 1 2 KPa For Pressure KPa me o o Temp C E E full RWT Kg s To account for the time elapsed after the break record the CASSIM iteration counts shown at the top right hand corner multiply that number by the time step 0 1 sec to get the time in seconds This calculation has assumed that the simulation iteration starts from 0 when the LOCA malfunction is initiated 117 METTUS EEE ere EE pa Pe ed 1 2 ROH 1 2 Temp P1 2 3 4 Flows to RIH s 1000 HTS process parameters at Pump Discharge 0 7 72213 2 20 11PM D 17 54 PM 2 20 11 PM ROH 1 ROH 2 Pressure 5 Feed Bleed Flow fd 0 ET bd
137. ve MSV outside containment leading to rapid depressurizaton of the main steam pressure Turbine generator will be runback rapidly and will be tripped by zero forward power The turbine trip initiates a reactor power setback The pipe break also results in increase in steam flow from the steam generators leading to increase in heat removal from the reactor coolant system Therefore coolant temperature and pressure will drop reactor coolant system screen insert the malfunction 100 main steam header break Observe and record the steam flows from the steam generators and the main steam pressure Observe the coolant temperature and pressure responses Observe that the turbine is running back to zero power Confirm turbine is tripped Record reactor power after setback Continue to monitor coolant pressure and temperature transients Discuss any safety margin challenge if any in this malfunction event and how the safety and control systems can cope with these challenges Reactor Trip _ Turbine Tr _ ROH Press Lo Lo Step Back Req d Setback Req d Turbine Runback Hi Neutron Pwr ROH Press Hi Hi Stm Gen Level Lo PRZR Lvl Hi Low Fwd Pwr Trip Main BFP s Trip Hi Neut Pwr LogR ROH Press Hi Main Stm Pres Hi Stm Gen Level Hi Loss RC Pmp s Main Steam Header P To HP heaters GENERATOR OUTPUT gross 0 00 MW 2 02 45 PM Steam CSDV ASDV MSSV Flow 1200 0 EZ
138. verage ROH Pressure Flow kg s BOP STM Flow ROSE aoe ee LT TT TS REENEN This screen shows the status of the shutdown system 1 SDS1 as well as the respective positions and average speed of the 18 zone control units ZCU Similarly the positions and the speeds of the two banks of absorber rods MCA are also displayed The reactivity contributions from the reactor feedback effects and each reactivity control device shutoff rods zone control units ZCU absorber rods MCA and gadolinium can be observed from the pop up window by pressing the Reactivity button on the top left of the screen All the reactivity devices considered for regulation as well as shutdown purposes are installed from above the Calandria known as the Reactivity Mechanism Deck Reactivity control is provided for the following effects 1 Long term bulk reactivity is mainly controlled by on power refuelling This is the only method for adding absolute positive reactivity to the core 2 Small frequent reactivity changes for both global and spatial neutron power are controlled by the zone control unit ZCU system 3 Negative reactivity to supplement the zone control units ZCU particularly for fast power reductions and to override the negative fuel temperature effect for large power level decreases is provided by the insertion of mechanical control absorbers MCA from their 37 normal poised position abov
139. y to such an extent that heat removal by normal means is not assured When the HTS pressure drops below the rupture pressure of the one way rupture discs the rupture discs burst thereby enabling emergency coolant injection to the RIH Water is injected into the heat transport system from the pressurized ECI accumulators Valves HHPV9 on the ECI interconnect line between the reactor outlet headers ROH open upon detection of a LOCA to assist in establishing a sustainable cooling flow path To enhance the effectiveness of the high pressure injection of water into the heat transport system the main steam safety valves MSSV1 to 4 open on detection of a LOCA to provide a rapid cool down of the steam generators and depressurization of the heat transport system High pressure injection continues until the ECI accumulators are nearly empty at which time the Long Term Cooling LTC system begins operation in long term recovery mode At this time the ECI injection valves MV1 MV2 close to ensure there is no injection of nitrogen gas into the HTS this is backed up by floating ball seals inside the ECI accumulators For a LOCA the LTC system is initiated during the operation of the ECI system On detecting a LOCA water is automatically introduced into the containment sumps and the LTC pumps start automatically When the water accumulators are nearly empty the ECI accumulator isolation valves close The recovery stage begins by pumping water from
140. ystem No 2 SDS2 SDS2 provides a second independent method of quickly terminating reactor power operation by injecting a strong neutron absorbing solution gadolinium nitrate into the moderator when any two out of three trip channels are tripped by any parameter c Emergency Core Cooling ECC System The ECC system is designed to supply water emergency coolant to the reactor core to cool the reactor fuel in the event of a loss of coolant accident LOCA The design bases events are LOCA events where ECC is required to fill and maintain the heat transport circuit inventory 17 Emergency Coolant Injection System 18 Schematic Diagram of the LTC System 19 d Containment System The containment system is a safety system with the function of limiting releases of radioactive material from within containment generally it prevents releases in excess of the site dose limits The containment system consists of the reactor building and liner electrical and process penetrations and other appurtenances which together form the containment envelope In addition the following subsystems act to ensure the continuity of the containment envelope or to reduce the contained pressures and energies following an accident a Main and auxiliary airlocks for the passage of personnel equipment and fuel b Containment isolation c Equipment for hydrogen control e g air coolers for mixing passive auto catalytic recombiners for H

(ACR-700) Simulator - International Atomic Energy Agency

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