Joint ICTP-IAEA Workshop on the Evaluation of Nuclear Joint - PowerPoint PPT Presentation

Joint ICTP-IAEA Workshop on the Evaluation of Nuclear Joint ICTP-IAEA Workshop on the Evaluation of Nuclear Reaction Data for Applications Reaction Data for Applications 2-13 Oct 2017, ICTP, Trieste, Italy 2-13 Oct 2017, ICTP, Trieste, Italy PCA benchmark analysis using the PCA benchmark analysis using the ADVANTG3.0.1/MCNP6.1.1b codes ADVANTG3.0.1/MCNP6.1.1b codes Mario Matijevi ć Mario Matijevi ć University of Zagreb, Faculty of Electrical Engineering and University of Zagreb, Faculty of Electrical Engineering and Computing, Department of Applied Physics Computing, Department of Applied Physics Unska 3, 10000 Zagreb, Croatia Unska 3, 10000 Zagreb, Croatia mario.matijevic@fer.hr

Contents: Contents: 1. PCA benchmark definition 2. PCA benchmark description 3. ADVANTG3.0.1/ MCNP6.1.1b codes 4. PCA response functions 5. PCA results using ADVANTG3.0.1/ MCNP6.1.1b 6. Summary and conclusions 2

1. PCA benchmark definition 1. PCA benchmark definition � Pool Critical Assembly Pressure Vessel (PCA) benchmark Pool Critical Assembly Pressure Vessel (PCA) benchmark � – A well known benchmark from the SINBAD database – A well known benchmark from the SINBAD database – Based on the PCA facility experiments at the ORNL – Based on the PCA facility experiments at the ORNL – A small pool- -type highly enriched experimental reactor type highly enriched experimental reactor – A small pool – – Measured and calculated (DORT) equivalent fission fluxes Measured and calculated (DORT) equivalent fission fluxes – – Validation of transport theory codes and XS libraries Validation of transport theory codes and XS libraries – – Prediction of the in- Prediction of the in -vessel neutron flux gradients vessel neutron flux gradients � � Scope of the PCA benchmark Scope of the PCA benchmark – Validation of the methodology to predict the reaction rates in ex x- -core region core region – Validation of the methodology to predict the reaction rates in e – Qualification of the pressure vessel fluence calculation methodology logy – Qualification of the pressure vessel fluence calculation methodo – Simulation of the neutron flux gradient inside carbon steel (RPV) ) – Simulation of the neutron flux gradient inside carbon steel (RPV – RPV surveillance programs of existing USA NPPs NPPs (RPV damage) – RPV surveillance programs of existing USA (RPV damage) – – Measured RR inside RPV (A4, A5, A6) and water gap in- Measured RR inside RPV (A4, A5, A6) and water gap in -front (A3, A2) front (A3, A2) – – Well defined neutron source, materials, and simple geometry Well defined neutron source, materials, and simple geometry – – DORT libraries in the PCA benchmark: BUGLE- DORT libraries in the PCA benchmark: BUGLE -93, SAILOR 93, SAILOR- -95, BUGLE 95, BUGLE- -96 96 – Computational requirements by the U.S. NRC Regulatory Guide 1.190 0 – Computational requirements by the U.S. NRC Regulatory Guide 1.19 3

2. PCA benchmark description 2. PCA benchmark description � PCA benchmark “12/13” configuration – PCA core with components mock up the core- -to to- -cavity region in cavity region in PWRs PWRs – PCA core with components mock up the core – Al plate, Thermal shield (TS), pressure vessel simulator (PVS), void box (VB) void box (VB) – Al plate, Thermal shield (TS), pressure vessel simulator (PVS), – Water gap between Al plate and TS: 12 cm – Water gap between Al plate and TS: 12 cm – – Water gap between TS and RPVS: 13 cm Water gap between TS and RPVS: 13 cm – – PCA facility is immersed in a large pool of water (coolant and moderator) PCA facility is immersed in a large pool of water (coolant and m oderator) – – PCA core PCA core has has 25 material test reactor (MTR) plate 25 material test reactor (MTR) plate- -type elements (e=93%) type elements (e=93%) MC model of PCA benchmark facility PCA pressure vessel wall benchmark facility 4 (water removed)

2. PCA benchmark description 2. PCA benchmark description � PCA benchmark “12/13” configuration Boron carbide Lead (1.6 g/cc) (11.34 g/cc) Al-U alloy (fuel plate) Aluminum PCA standard MTR fuel element (water included) PCA control rod (water included) Cross sectional view through the control element 5

2. PCA benchmark description 2. PCA benchmark description � PCA benchmark “12/13” configuration 6

2. PCA benchmark description 2. PCA benchmark description � PCA benchmark “12/13” configuration 7

2. PCA benchmark description 2. PCA benchmark description � Experimental measurements: A1 – A7 Equivalent fission fluxes at detector locations: reaction rates φ = σ eq i = ∫ σ φ ( ) ( ) E E dE i φ E ∫ eq σ ϕ ( ) ( ) E E dE i E ∫ ϕ ( ) E dE E Maxwell 1/ E slowing fission 8

2. PCA benchmark description 2. PCA benchmark description � Reactions of interest ( Reactions of interest ( shielding calculations ) � shielding calculations ) – 237 Np(n,f) 237 Np(n,f) 137 137 Cs, Cs, 238 238 U(n,f) U(n,f) 137 137 Cs, Cs, 103 103 Rh(n,n') Rh(n,n') 130m 130m Rh, Rh, 115 115 In(n,n') In(n,n') 115m 115m In, In, 58 58 Ni(n,p) Ni(n,p) 58 58 Co, Co, – 27 Al(n, 24 Na 27 Al(n, α α ) ) 24 Na – – Results are given per unit PCA core neutron source (normalized) Results are given per unit PCA core neutron source (normalized) 235 U fission fluxes – Calculated- -to to- -measured (C/M) ratios of equivalent measured (C/M) ratios of equivalent 235 U fission fluxes – Calculated – Measurements in core midplane (z=0 and y=0) at at locations A1 to A7 – Measurements in core midplane (z=0 and y=0) locations A1 to A7 – Experimental access tubes: steel in PV and Plexiglas in water locations cations – Experimental access tubes: steel in PV and Plexiglas in water lo – Minimization of the perturbations of the neutron field – Minimization of the perturbations of the neutron field – DORT reaction rates (/atom/s) are given for dosimeters A1- -A8 A8 – DORT reaction rates (/atom/s) are given for dosimeters A1 � Critical configuration ( Critical configuration ( eigenvalue calculations ) � eigenvalue calculations ) – – Fully inserted control rod reaches bottom of the fuel Fully inserted control rod reaches bottom of the fuel – – Withdrawn length is measured from the bottom of the fuel Withdrawn length is measured from the bottom of the fuel – Safety rods (S1, S2 and S3) critical positions: 48.26 cm – Safety rods (S1, S2 and S3) critical positions: 48.26 cm – Regulating rod (RR) position: 38.43 cm – Regulating rod (RR) position: 38.43 cm – Total critical mass of 235 235 U: 3336.01 g U: 3336.01 g – Total critical mass of 9

3. ADVANTG3.0.1/ MCNP6.1.1B codes 3. ADVANTG3.0.1/ MCNP6.1.1B codes � ADVANTG3.0.1 code (ORNL) ADVANTG3.0.1 code (ORNL) � – Automated mesh- -based tool for generating VR parameters for MCNP code based tool for generating VR parameters for MCNP code – Automated mesh – Approximate 3- -D multigroup SN forward/adjoint transport solutions D multigroup SN forward/adjoint transport solutions – Approximate 3 – Denovo SN solver developed at ORNL with CADIS/FW- -CADIS formalism CADIS formalism – Denovo SN solver developed at ORNL with CADIS/FW – – VR parameters: space- VR parameters: space -energy weight energy weight- -windows (WW) and biased source distributions windows (WW) and biased source distributions (SB) cards for the (SB) cards for the MCNP input MCNP input � MCNP6.1.1b code (LLNL) � MCNP6.1.1b code (LLNL) – general- -purpose Monte Carlo N purpose Monte Carlo N- -Particle code with arbitrary 3D geometry Particle code with arbitrary 3D geometry – general – neutron, photon, electron, or coupled n/p/e n/p/e transport – neutron, photon, electron, or coupled transport – XS libraries: continuous, discrete, multigroup, S(a,b S(a,b) law, dosimetry, ) law, dosimetry,… … – XS libraries: continuous, discrete, multigroup, – powerful general source, rich collection of VR techniques, flexible tallies ble tallies – powerful general source, rich collection of VR techniques, flexi – – Pointwise XS data with MAKXSF for XS libraries with Doppler broadening Pointwise XS data with MAKXSF for XS libraries with Doppler broa dening PCA model with MCNP in z=0 cm plane PCA model with MCNP in y=30.84 cm plane 10

3. ADVANTG3.0.1/ MCNP6.1.1B codes 3. ADVANTG3.0.1/ MCNP6.1.1B codes Hybrid shielding � � forward Adjoint forward flux = σ † ( , q r E ) ( , r E ) transport: d source source � � � 1 = φ † q r E ˆ( , ) q r E ( , ) ( , r E ) Biased detector R source � R CADIS = Target � w r E ( , ) φ † weight ( , r E ) � � q r E ( , ) R Initial = = � � w r E ( , ) 0 φ † weight q r E ˆ( , ) ( , r E ) adjoint flux adjoint � transport: CADIS σ � FW- detector ( , r E ) Weighted = † d q r E ( , ) � � adjoint ∫ source σ φ ( , r E ) ( , r E ) source d point detector = adjoint source detector source MCNP model of the ¼ cask ADVANTG SN mesh for VR parameters 11 Adjoint function (1.8-2.4) MeV