Neutronics, Thermal Hydraulics and Safety Parameter Studies of the 3 MW TRIGA Research Reactor at AERE, Savar

Transcription

1 Neutronics, Thermal Hydraulics and Safety Parameter Studies of the 3 MW TRIGA Research Reactor at AERE, Savar Md. Quamrul HUDA Energy Institute Atomic Energy Research Establishment Bangladesh Atomic Energy Commission Savar, Dhaka.

2 The 3 MW TRIGA Mk-II Research Reactor of Bangladesh Atomic Energy Commission TRIGA: Training Research Isotope production General Atomics Shield Structure of the TRIGA Reactor.

3 Cutaway view of TRIGA reactor.

4 Cutaway view of TRIGA reactor.

5 Final core arrangement of the TRIGA reactor.

6 TRIGA stainless-steel-clad fuel element with triflute and fittings.

7 Fuel-follower control rod withdrawn and inserted position.

8 OUTLINE Neutronic Analysis of the TRIGA Reactor Generation of Cross Section Library Effective Multiplication Factor Neutron Flux Distribution Power Distribution and Power Peaking Factors Control Rod Worth Critical Rod Height and Excess Reactivity Shutdown Margin

9 OUTLINE Thermal Hydraulic Analysis of TRIGA Reactor Analysis by NCTRIGA for Natural Convection Steady-State Thermal Hydraulic Analysis Temperature Distribution Effect of Coolant Mass Flow Rate DNB Analysis Thermal Hydraulic Analysis for the Transients Loss-of-Flow Accident Analysis Accidental Reactivity Insertion Loss-of Coolant Accident Analysis Concluding Remarks

10 Computer Codes Used Neutronics: NJOY99.0 MCNP4C WIMS-D5/CITATION Thermal Hydraulics: NCTRIGA PARET RELAP5 Accident Analysis: PARET RELAP5

11 Flow chart for coupled neutronics/thermal hydraulic analysis.

12 MCNP modeling of the TRIGA MARK II research reactor.

13 Cutaway view of TRIGA active core comprising rotary specimen rack Lazy Susan and beam ports.

14 Data Library Used: ENDF/B-VI & Jendl3.3 for 235 U, 238 U, 1 H, 166 Er, 167 Er, 12 C, 55 Mn, 10 B, 11 B, 16 O, 27 Al, 207 Pb ENDF/B-V & Jendl3.2 for nat Zr, nat Fe, nat Cr, nat Mo, nat Ni, nat Mg, nat Si ENDF/B-V for S(a,b)

15 Flow chart for generating continuous energy cross section data.

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17 Cutaway view of TRIGA active core comprising rotary specimen rack Lazy Susan and beam ports.

18 Cutaway view of TRIGA active core comprising rotary specimen rack Lazy Susan and beam ports.

19 Results from Neutronic Analyses

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22 Comparison between the measured and MCNP calculated axial flux distribution at 3 MW along the water filled central thimble and rotary specimen rack Lazy Susan, respectively.

23 Comparison between the measured and MCNP calculated axial saturated activity distribution at 3 MW along the water filled central thimble and rotary specimen rack Lazy Susan, respectively.

24 Neutron flux distribution within the DCT

25 Power distribution within the fuel and fuel-follower elements at 3MW.

26 Axial and radial power distribution within the TRIGA fuel elements.

27 Comparison between the experiment and MCNP calculated control rod worths for different control elements.

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29 Thermal Hydraulic Analysis

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31 NCTRIGA modeling of TRIGA fuel.

32 PARET modeling of TRIGA fuel.

33 Comparison of peak thermocouple temperatures measured in C1 and D2 IFEs under steady-state condition at different power levels and modeled in PARET.

34 TRIGA instrumented fuel element.

35 Ratio of the peak fuel temperature calculated in C1 and D2 IFEs using PARET to the measured peak thermocouple temperature under steady-state condition.

36 Axial temperature and fuel surface heat flux & heat transfer coefficient calculated by PARET in the hottest fuel channel (C4) under steady-state condition at full power.

37 Variation of temperatures at the center of the hottest fuel element and its associated coolant channel at varying flow rates modeled in PARET.

38 The design conditions used for DNB calculation by PARET Operating Power 3MW Flow rate kg/m 2 -s (3500 gpm) Inlet temperature o C (105 o F) Pressure X10 2 kpa Hot Rod Factor (HFR) (C4 element) Inlet pressure loss coefficient 1.81 Outlet pressure loss coefficient 2.12 Qualitative representation of heat flux and related conditions along the hot channel in the 3 MW TRIGA reactor.

39 Fuel performance at different power levels of the TRIGA reactor. Fuel performance at various flow rates.

40 Fuel performance at different hot rod factors. Fuel performance at different inlet temperatures.

41 Flow coast down transient response of TRIGA core to a loss-of-coolant flow with a decay time of 25.0 sec, a scram trip at 85% and a sec delay.

42 Flow coast down temperatures at the fuel centerline, clad and coolant exit after the transient response of the TRIGA core to loss-of-flow coolant accidents for the hot and average channel.

43 RELAP5 input model for TRIGA

44 Steady-state analysis of TRIGA using RELAP5 at full power Fuel centerline temperature: 1. ~750 o C (Expt.) 2. ~790 o C (RELAP5) CORE FLOW

45 Features included: Fast Loss of flow analysis 1. Pumps were replaced by branch and time-dependent junction 2. Control variable (51) was included to to reduce the flow as e -t/t with T=1 sec 3. Variable trips were set at 85% of nominal flow with 200 ms delay 4. Logical trip was included to scram the reactor thro the linear reactivity insertion of $ (total control rod worth) in 0.5 sec EXPONENTIAL LOSS OF FLOW

46 Fast Loss of flow analysis CORE POWER / REACTIVITY LOSS OF FLOW

47 Features included: Fast Reactivity Insertion Analysis 1. Ramp reactivity insertion of $1.50 in 0.5 sec at 1 watt reactor power 2. Safety system trip point was set at 1.1P o (3.3 MW) 3. Variable trips were used to insert the ramp with 25 ms delay before the control rods insertion is initiated INSERTED REACTIVITY

48 C O R E Fast Reactivity Insertion Analysis P O W E R & F L O W

49 Beam Tube Rupture Loss of Coolant Accident

50 D R A I N E D F L O W Beam Tube Rupture Loss of Coolant Accident MIXING IS GOING ON IN THE REACTOR TANK

51 P O W E R Beam Tube Rupture Loss of Coolant Accident & S C R A M Scram was set to 25% of voidf in reactor tank C O R E F L O W

52 Beam Tube Rupture Loss of Coolant Accident V O L U E M E T R I C P R E S S U R E

53 CONCLUDING REMARKS MCNP has been used to develop a versatile and accurate reactor physics model of the TRIGA reactor. Continuous energy cross-section data from ENDF/B-VI & JENDL-3.3 in combination with ENDF/B-V & JENDL-3.2 and S(a,b) scattering functions from the ENDF/B-V library were used. Most of the TRIGA reactor benchmark experiments were simulated to validate the physical model. Excellent agreement between the experiment and the MCNP calculation indicates that the MCNP simulations might be used as reference with confidence for further core configuration studies. The MCNP model of the TRIGA is one of very few low-enrichment benchmarks available and is providing a valuable resource to validate evaluated data files. Its continued use will greatly expedite future research studies, especially to provide feedback to the neutronic evaluation of the burnt core calculation. It is found that the MCNP4C - NCTRIGA - PARET codes can be used effectively for the thermal hydraulic analysis of the TRIGA research reactor both under natural- and forced-convection mode of coolant flow to predict the safety parameters. Contd.

54 CONCLUDING REMARKS The RELAP/SCDAPSIM/MOD3.4 computer code has been used to successfully perform accident analyses to support the Safety Analysis of the 3.0 MW TRIGA research reactor. RELAP5 has been used to develop a versatile and realistic model of the TRIGA research reactor. However, it requires some more refinement about the TRIGA parameters before making any exact comparison, but the primary results are encouraging and indicate that the RELAP5 might be used as reference with confidence for TRIGA thermal hydraulic analysis. Future collaboration is required between the RELAP5 user communities to share knowledge and experience. The RELAP model of the TRIGA is one of very few research reactor models available that may provide a valuable resource to validate the RELAP5 code for research reactor analysis. Its continued use will greatly expedite future research studies, especially to get feedback regarding safety aspects of nuclear reactors.

55 REFERENCES S.I. Bhuiyan, M.M. Sarker, M. Rahman, M.S. Shahdatullah, M.Q. Huda, T.K. Chakrobortty, M.J.H. Khan, Criticality and Safety Parameter Studies of a 3-MW TRIGA MARK II Research Reactor and Validation of the Generated Cross-section Library and Computational Method. Nucl. Technol. 130, M.Q. Huda, M. Rahman, M.M. Sarker and S.I. Bhuiyan, Benchmark Analysis of the TRIGA MARK II Research Reactor using Monte Carlo Techniques, Annals of Nuclear Energy, 31(11): (2004). M.Q. Huda, S.I. Bhuiyan, T.K. Chakrobortty, M.M. Sarker and M.A.W. Mondal, Thermal Hydraulic Analysis of the 3 MW TRIGA MARK II Research Reactor, Nuclear Technology, 135 (1), (2001). M.Q. Huda and M. Rahman, Thermo-hydrodynamic Design and Safety Parameter Studies of the TRIGA MARK II Research Reactor, Annals of Nuclear Energy, 31(10): (2004). M.Q. Huda and S.I. Bhuiyan, Investigation of Thermohydraulic Parameters during Natural Convection Cooling of TRIGA Reactor, Annals of Nuclear Energy, 33(13): (2006). A. R. Antariksawan, M. Q. Huda, Tiancai Liu, J. Zmitkova, C. M. Allison and, J. K. Hohorst, Validation of RELAP/SCDAPSIM/MOD3.4 for Research Reactor Applications, Proc. Intl. Conf. on Nucl. Eng., ICONE , Beijing, China, May 16-20, 2005.

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