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Transient Analysis of Primary Feed Pump Trip for 700 Mwe IPHWR

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Transient Analysis of Primary Feed Pump Trip for 700 Mwe IPHWR Kansas State University Libraries New Prairie Press Symposium on Advanced Sensors and Modeling Techniques for Nuclear Reactor Safety Transient Analysis of Primary Feed Pump Trip for 700 MWe IPHWR S. Phani Krishna Nuclear Power Corporation of India, [email protected] S. Pahari Nuclear Power Corporation of India, [email protected] S. Hajela Nuclear Power Corporation of India, [email protected] M. Singhal Nuclear Power Corporation of India, [email protected] See next page for additional authors Follow this and additional works at: https://newprairiepress.org/asemot Part of the Mechanical Engineering Commons, and the Nuclear Engineering Commons Recommended Citation Krishna, S. Phani; Pahari, S.; Hajela, S.; and Singhal, M. (2018). "Transient Analysis of Primary Feed Pump Trip for 700 MWe IPHWR," Symposium on Advanced Sensors and Modeling Techniques for Nuclear Reactor Safety. https://newprairiepress.org/asemot/2018/fullprogram/19 This Contributed is brought to you for free and open access by the Conferences at New Prairie Press. It has been accepted for inclusion in Symposium on Advanced Sensors and Modeling Techniques for Nuclear Reactor Safety by an authorized administrator of New Prairie Press. For more information, please contact [email protected]. Presenter Information S. Phani Krishna, S. Pahari, S. Hajela, and M. Singhal This contributed is available at New Prairie Press: https://newprairiepress.org/asemot/2018/fullprogram/19 TRANSIENT ANALYSIS OF PRIMARY FEED PUMP TRIP FOR 700 MWe INDIAN PRESSURIZED HEAVY WATER REACTOR S. Phani Krishna*, S. Pahari, S. Hajela, M. Singhal Nuclear Power Corporation Of India Limited, NUB, Anushakti Nagar, Mumbai, PIN:-400094, *email: [email protected] ABSTRACT METHODOLGY/FORMULATION 700 MWe Indian Pressurized Heavy Water Reactor NODALIZATION OF ATMIKA.T: (IPHWR) is a horizontal channel type reactor with two loops of Primary Heat Transport (PHTS) system. Three (two operating and one stand by) main boiler feed water pumps (BFP) supply feed water to Steam Generators (SGs). In the event of one of the running BFP trip, standby comes on line on auto. The transient has been initiated by tripping one of the pumps. In this paper two cases are considered for analysis with postulated initiating event as follows: Case 1: BFP Trip and Standby BFP available on auto Case 2: BFP Trip and Standby pump not available. • Fluid Dynamics Model • Drift Flux Model Transient analysis was performed using ATMIKA-T in- • Flow Stratification house developed 3-D neutron kinetics coupled thermal • Critical Discharge Model Thermal • Frictional Model for 1-Φ & hydraulics computer code. This paper provides timeline of Hydraulic Model 2-Φ region • Heat Conduction sequence of events which is important for operator’s • Wall Heat Transfer Mode • Metal Water Reaction action to maneuver the event. • Core Neutronics Model • Reactor Regulating System PRIMARY FOCUS/OBJECTIVES OF Sequences of events in case 1 and case 2: • Steam Generator Pressure/Level Control (SGPC/SGLC) • Feed Flow reduces. • PHT Pressure Control THE STUDY • Pressurizer Level Control : Control System Model • BCD Pressure/Level Control • SGs Pressure Rises due to reduction in sub-cooled • Electro Hydraulic Turbine Control The objective of this paper is to analyse/study plant (EHTC) • Various Control related Interlocks dynamics during trip of Primary feed pump trip during water availability and non availability of stand by feed pump • SGs level starts falling. If standby comes, it recovers of 700 MWe IPHWR and bring out the sequence of events and reactor stabilizes at 100% FP. • Pumps which is important for deciding operator’s action to • Valves maneuver the event. • If Stand by fails to start then • Pressurizer • Accumulator • Reactor set back on SG low level Component • Fuel Pin Model • Pipe • Reactor trips on SG level very very low • Heat Exchanger RESULTS • Turbine Trips 100 1.20 2 4 4 ) case i ) ) g ) case ii ( 2 g ) ( s 2 m / 0 ) 0 c g m 0 / m c k 0 g / ( ( k l g e ( t k case i P 0.80 e ( a v e F case i case i case i r case ii e e -4 R f l -2 u r case ii o case ii s case ii u w G s s n o l s e -100 S o -4 r i e F t r n P i c n P i -8 a e -4 r G n g e i F S 0.40 n g e n a n i g h a -200 n e -8 h C a -12 g C -6 h n C a h C -16 -300 0.00 -8 -12 -10 290 590 890 1190 1490 -10 290 590 890 1190 1490 -10 290 590 890 1190 1490 -10 290 590 890 1190 1490 -10 290 590 890 1190 1490 Time(Seconds) Time(Seconds) Time(Seconds) Time(Seconds) Time(Seconds) CHANGE IN FEED FLOW WRT THE SG LEVEL VARIATION WRT SG PRESSURE VARIATION WRT PHTS PRESSURE VARIATION WRT NOMINAL FLOW REACTOR POWER INITIAL LEVEL THE SET POINT THE SET POINT Case 1 Case 2 Event calculated calculated timeline (s) timeline (s) CONCLUSION Key References/ Feed Flow • In case 1, If the standby pump comes after 14 sec, the SG level recovers with a slight dip in level reduces 0 0 (does not reach the low level alarm setting). The feed flow rises and settles down to normal value. Acknowledgment Stand By Subsequently all the parameters converge to steady state value. Reactor continues to operate at comes after 14 - Reactor set 100% FP. [1] Rammohan H.P., and Bajaj S. back on SG S., 1996. “ATMIKA-A System level low - 365 • In case 2 without the availability of standby BFP, feed flow rate drops. Steam Generator pressure rises slightly due to reduction in sub cooled feed flow and SG level starts to decrease. Reactor Thermal-Hydraulic Computer Reactor Trips on SG level setback starts as SG level goes below setback setting at around 6 min. SG level continues to fall Code Model Description”., - 450 very very low and reactor trips on SG Level very very low at around 7.5 min. Thereafter, core decay heat removal Internal Report, NPCIL, continues initially by PCPs coast down and subsequently by thermosyphoning on primary side. Mumbai, India, 1996. TIME LINE OF EVENTS • For both the cases, automatic actions are sufficient to safely ride over the postulated transient. "Advanced Sensors and Modeling Techniques for Nuclear Reactor Safety,December 15-19, 2018, IIT Bombay, Mumbai, India".
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