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Assessement of the JEFF-3.1.1 Neutron Data Library for FOR CSE of LWR Fuel Storage Pools
Edwin Kolbe,Alexander Vasiliev,Hakim Ferroukhi 한국물리학회 2011 THE JOURNAL OF THE KOREAN PHYSICAL SOCIETY Vol.59 No.23
A methodology for criticality safety evaluations (CSE) of light water reactor (LWR) compact storage pools and transport casks based on the Monte Carlo code MCNPX-2.5.0 was recently established at PSI. Validation calculations were performed applying a suite of 15 low-enriched thermal compound uranium and 4 mixed plutonium uranium thermal compound benchmarks extracted from the International Handbook of Evaluated Criticality Safety Benchmark Experiments. In a first part of this paper, the same validation suite, comprising in total 149 benchmark cases, is analyzed with the latest release of the JEFF-3.1.1 nuclear data library. The resulting bias is compared to the ones obtained previously with the ENDF/B-VII.0, JEFF-3.1 and JENDL-3.3 libraries. Secondly, in licensing-related studies for a new PWR commercial wet storage pool, a noticeable sensitivity upon the employed thermal neutron scattering S(α,β) matrix data was observed for water reflected configurations. By performing trend analyses with respect to the magnitude of the thermal flux for a selected subset of benchmark cases, various parameterizations of S(α,β) could be assessed including the one in JEFF-3.1.1. Finally, the benefits from applying several cross-section libraries in CSE will be pointed out.
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Roman Mukin,Ivor Clifford,Omar Zerkak,Hakim Ferroukhi 한국원자력학회 2018 Nuclear Engineering and Technology Vol.50 No.3
A series of tests dedicated to station blackout (SBO) accident scenarios have been recently performed at thePrim€arkreislauf-Versuchsanlage (primary coolant loop test facility; PKL) facility in the framework of theOECD/NEAPKL-3 project. These investigations address current safety issues related to beyond design basis accidenttransients with significant core heat up. This work presents a detailed analysis using the best estimatethermalehydraulic code TRACE (v5.0 Patch4) of different SBO scenarios conducted at the PKL facility; failuresof high- and low-pressure safety injection systems together with steam generator (SG) feedwatersupply are considered, thus calling for adequate accidentmanagement actions and timelyimplementation ofalternative emergency cooling procedures to prevent core meltdown. The presented analysis evaluates thecapability of the applied TRACE model of the PKL facility to correctly capture the sequences of events in thedifferent SBO scenarios, namely the SBO tests H2.1, H2.2 run 1 and H2.2 run 2, including symmetric orasymmetric secondary side depressurization, primary side depressurization, accumulator (ACC) injection inthe cold legs and secondary side feeding with mobile pump and/or primary side emergency core coolantinjection from the fuel pool cooling pump. This study is focused specifically on the prediction of the core exittemperature, which drives the execution of the most relevant accident management actions. This workpresents, in particular, the key improvements made to the TRACE model that helped to improve the codepredictions, including the modeling of dynamical heat losses, the nodalization of SGs' heat exchanger tubesand theACCs. Another relevant aspect of thiswork is to evaluate howwell themodel simulations of the threedifferent scenarios qualitatively and quantitatively capture the trends and results exhibited by the actualexperiments. For instance, howthenumber of SGs considered for secondary side depressurization affects theheat transfer from primary side; how the discharge capacity of the pressurizer relief valve affects the dynamicsof the transient; how ACC initial pressure and nitrogen release affect the grace time between ACCinjection and subsequent core heat up; and howwell the alternative feeding modes of the secondary and/orprimary side with mobile injection pumps affect core quenching and ensure stable long-term core coolingunder controlled boiling conditions.
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