Sanchez-Espinoza, V.

Schlenker, M.

English

KITopen

KIT – University of the State of Baden-Wuerttemberg and  National Research Center of the Helmholtz Association Institute for Neutron Physics and Reactor Technology www.kit.edu Increasing the Fuel Utilization in Gen-II BWR with Reduced-Moderation Square Lattice Fuel Assemblies  ICAPP2014-14145 Markus Schlenker  &  Victor Sanchez-Espinoza     Karlsruhe Institute of Technology,                                               Institute for Neutron Physics and Reactor Technology         2014 International Congress on Advances in Nuclear Power Plants (ICAPP)  Westin Charlotte, North Carolina, U.S.A. April 6-9, 2014 Institute for Neutron Physics and Reactor Technology 2 Motivation Improve fuel utilization in light water reactors (LWR) by increasing the conversion ratio  Make multi-recycling more feasible  Use advantages of  boiling water reactor (BWR)  Maintain inherent safe behavior  Use proven plant design (Gen-II BWR: German KWU series 72)  04.04.2014 Markus Schlenker ICAPP2014 - 14145 Introduction Design studies Dancoff factors Validation studies Conclusions Institute for Neutron Physics and Reactor Technology 3 Solution approach 04.04.2014 Markus Schlenker ICAPP2014 - 14145 Iterations Fuel assembly (FA) based analysis with lattice physics code: 3-D core analysis (not in this paper):  Assess global safety behavior  Obtain average T-H conditions for more representative depletion analysis (FA level) ATRIUMTM 10XM (reference) Low moderated FA Reduce Moderation:  Remove water structures  Modify fuel pin radius  Modify pitch  Keep FA-pitch, FA-channel, control rod geometry Validation with Monte Carlo code:  Confirm deterministic solution for harder neutron spectrum  Confirm void reactivity coefficient Introduction Design studies Dancoff factors Validation studies Conclusions Institute for Neutron Physics and Reactor Technology 4 SCALE6.1 code package TRITON lattice physics sequence: ENDF/B-VII multigroup cross- sections in 238 energy groups BONAMI/CENTRM/PMC for multi- group cross-section processing Manual definition of Dancoff factors NEWT deterministic 2-D SN code Flexible mesh ORIGEN-S depletion code  KENO-VI Monte Carlo code ENDF/B-VII cross-sections   continuous energy  multigroup (238 energy groups) BONAMI/CENTRM/PMC for multigroup cross-section processing   04.04.2014 Markus Schlenker ICAPP2014 - 14145 M. D. DeHart, S. M. Bowman 2010, Reactor Physics Methods and Analysis Capabilities in Scale, Nuclear Technology, Vol. 174, May 2011 Introduction Design studies Dancoff factors Validation studies Conclusions Institute for Neutron Physics and Reactor Technology 5 Parametrization:         Fuel material compositions: Parametric design studies (TRITON / NEWT)  04.04.2014 Markus Schlenker ICAPP2014 - 14145 Case 1 2 3 4 5 6 7 Ref Lattice dimension 10x10 10x10 10x10 12x12 10x10 10x10 10x10 10x10-9 Pitch, cm 1.290 1.285 1.285 1.070 1.285 1.285 1.285 - Rrod, cm 0.545 0.573 0.573 0.465 0.573 0.573 0.573 0.514 Rfuel, cm 0.470 0.494 0.494 0.401 0.494 0.494 0.494 0.452 Rod distance, cm 0.20 0.14 0.14 0.14 0.14 0.14 0.14 - moderator to fuel ratio (MFR) 1.842 1.543 1.543 1.695 1.543 1.543 1.543 2.477 Fuel volume*, cm³/cm 69.5 76.7 76.7 72.8 76.7 76.7 76.7 58.4 Av. enrichm., wt-% 5 5 6 5 5 5 6 4.4 Pu-Vector (see below) 1 1 1 1 2 3 3 1 Pu-Vector Pu238 Pu239 Pu240 Pu241 Pu242 Pufiss 1 2 56.5 26.1 8.6 6.8 65.1 2 4 48 31 7 10 55 3 4 38 33 12 13 50 *Minimum rod clearance in study * Introduction Design studies Dancoff factors Validation studies Conclusions Institute for Neutron Physics and Reactor Technology 6 Results for keff and CR (5% Pufiss) 04.04.2014 Markus Schlenker ICAPP2014 - 14145 Case 1 2 4 5 6 Ref Lattice dimension 10x10 10x10 12x12 10x10 10x10 10x10-9 Rrod, cm 0.545 0.573 0.465 0.573 0.573 0.514 moderator to fuel ratio (MFR) 1.842 1.543 1.543 1.695 1.543 1.543 1.543 2.477 Pu-Vector (see below) 1 1 1 2 3 1 Introduction Design studies Dancoff factors Validation studies Conclusions Keff decrease slower due to higher CR Results not directly comparable due to differing achievable cycle length   Institute for Neutron Physics and Reactor Technology 7 Apply linear reactivity model (Driscoll, 1990) Assume leakage of ~2.5 % in effective system Matching cycle length for matching kinf at k=1.025 Example:         Interpolation suggest ~5.6% Pufiss for comparable cycle length Interpolating results to compare cases 04.04.2014 Markus Schlenker ICAPP2014 - 14145 k=1.025 k=1.025 with 6% Pufiss k=1.025 with 5% Pufiss Introduction Dancoff factors Validation studies Conclusions Design studies Institute for Neutron Physics and Reactor Technology 8 Results – Fuel utilization  Parametrization with adjusted enrichment for comparable cycle length           Slightly better fuel utilization with thicker rods (smaller MFR!) Best fuel utilization with lowest fresh Pu-quality Degradation of Pu-vector significantly reduced   04.04.2014 Markus Schlenker ICAPP2014 - 14145 Case 1 2 4* 5* 6* Ref Lattice dimension 10x10 10x10 12x12 10x10 10x10 10x10-9 Pitch, cm 1.290 1.285 1.070 1.285 1.285 - Rrod, cm 0.545 0.573 0.465 0.573 0.573 0.514 Rod distance, cm 0.20 0.14 0.14 0.14 0.14 - MFR 1.842 1.543 1.695 1.543 1.543 2.477 Av. enrichm., wt-% 5 5 5.3 5.6 5.6 4.4 Fresh Puqual 0.65 0.65 0.65 0.55 0.50 0.65 Cycle length, efpd 270 271 273 265 265 276 Cycle average CR 0.729 0.753 0.744 0.780 0.794 0.68 Pu-quality-change (discharge - fresh) -0.08 -0.06 -0.06 -0.03 -0.03 -0.16 Introduction Dancoff factors Validation studies Conclusions Design studies Institute for Neutron Physics and Reactor Technology 9 Results – Void reactivity coefficient 04.04.2014 Markus Schlenker ICAPP2014 - 14145 Case 1 2 3 4 5 6 7 Ref Lattice dimension 10x10 10x10 10x10 12x12 10x10 10x10 10x10 10x10-9 moderator to fuel ratio (MFR) 1.842 1.543 1.543 1.695 1.543 1.543 1.543 2.477 Av. enrichm., wt-% 5 5 6 5 5 5 6 4.4 Pu-Vector 1 1 1 1 2 3 3 1 Introduction Dancoff factors Validation studies Conclusions Design studies VC reduced but negative Case 1 2 3 4 5 6 7 Approximate limiting av. enr. for negative VC, wt-% ~8.7 ~7.7 ~7 ~6.7 ~6 Institute for Neutron Physics and Reactor Technology 10 User-defined Dancoff factors (C) Only needed for multi-group calculations Standard-approach: Infinite lattice C in every unit cell BWR: Quasi infinitive lattice in center → infinite lattice C ok Very heteregeneous peripheral lattice  → infinite lattice C not correct! C is void dependent C can be grouped (center, side, corner) e.g. in upper left quadrant of low moderated FA: 04.04.2014 Markus Schlenker ICAPP2014 - 14145 0 % void content 40 % void content 0.23 0.33 0.34 0.33 0.34 0.27 0.40 0.41 0.40 0.41 0.34 0.48 0.48 0.49 0.48 0.40 0.58 0.59 0.60 0.59 0.34 0.48 0.48 0.49 0.49 0.41 0.59 0.59 0.60 0.60 0.34 0.48 0.48 0.48 0.48 0.41 0.59 0.59 0.59 0.59 0.34 0.48 0.49 0.49 0.49 0.41 0.59 0.60 0.60 0.60 80 % void content 100 % void content 0.34 0.50 0.51 0.51 0.51 0.38 0.57 0.59 0.58 0.59 0.50 0.73 0.74 0.75 0.74 0.57 0.84 0.85 0.86 0.85 0.51 0.74 0.75 0.76 0.76 0.58 0.85 0.87 0.87 0.87 0.51 0.75 0.76 0.76 0.76 0.58 0.86 0.87 0.87 0.87 0.52 0.75 0.76 0.76 0.76 0.59 0.86 0.87 0.87 0.87 Introduction Design studies Dancoff factors Validation studies Conclusions Institute for Neutron Physics and Reactor Technology 11 Influence of C on results (NEWT) Overprediction by several 100 pcm without correct Dancoff factors Difference increases with void content Especially high disagreement for high void content 04.04.2014 Markus Schlenker ICAPP2014 - 14145 Introduction Design studies Dancoff factors Validation studies Conclusions Institute for Neutron Physics and Reactor Technology 12 Validation studies (NEWT vs. KENO-VI) Good agreement of NEWT and MG-KENO Difference of <1% between MG-KENO and CE-KENO Bad agreement for void > 80% Potential sources for differences between MG-KENO and CE-KENO : S(α,β) treatment in CE-KENO, Dancoff factors, Multi-group approximations 04.04.2014 Markus Schlenker ICAPP2014 - 14145 Introduction Design studies Dancoff factors Validation studies Conclusions Institute for Neutron Physics and Reactor Technology 13 „…the differences between the multigroup and continuous energy results are expected to be minimized with a pending S(α,β) update for the continuous energy cross sections that will be available with a subsequent release of Scale.”*  (Scale6.2) 04.04.2014 Markus Schlenker ICAPP2014 - 14145 Reference to KENO validation* *W. J. Marshall, B. T. Rearden 2011, Criticality Safety Validation of Scale 6.1,  ORNL/TM-2011/450  * MOX systems with  increasing moderation  (left to right) Considering of all approximations and KENO validation shows, 1% difference between MG-KENO and CE-KENO is reasonable  Prediction improvement with SCALE6.2 assessed in the future Introduction Design studies Dancoff factors Validation studies Conclusions Institute for Neutron Physics and Reactor Technology 14 Conclusions Design studies for low moderated FA: Improvement of the conversion ratio:   0.68  →  0.73 to  0.79 Slower degradation of plutonium quality: -0.16  → -0.03 to -0.08 → Second recycling more feasible Reduced but negative void reactivity coefficient Corrected Dancoff factors improve result significantly  Validation of NEWT model with KENO shows reasonable agreement Potential sources for differences between MG-KENO and CE-KENO: S(α,β) treatment in CE-KENO Dancoff factors Multi-group approximations High void content predictions disagree (KENO vs NEWT)  and have to be investigated in the future  Core calculations are needed for more representative BWR conditions 04.04.2014 Markus Schlenker ICAPP2014 - 14145 Introduction Design studies Dancoff factors Validation studies Conclusions Institute for Neutron Physics and Reactor Technology 15 Acknowledgements The authors gratefully acknowledge the technical advice of Prof. H.-D. Berger, Dr. M. Rost and all other involved colleagues of AREVA GmbH and want to thank for the financial support of AREVA GmbH for this work.  This work is also supported by the Reactor Safety Research Program of the Karlsruhe Institute of Technology (KIT). 04.04.2014 Markus Schlenker ICAPP2014 - 14145 

Inceasing the fuel utilization in Gen-II BWR with reduced-moderation square lattice fuel assemblies

https://publikationen.bibliothek.kit.edu/240095876/3816547

Inceasing the fuel utilization in Gen-II BWR with reduced-moderation square lattice fuel assemblies

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