Abstract
The topic addressed deals with the determination of adjoint parameters for instrumentation relevance. This is a crucial subject for comprehension of subcritical levels in the frame of safety analysis. Indeed, such states require interpretation and raw data cannot be processed as such. To do so, the transcription of core reactivity through instrumentation located in the reactor periphery is considered with the use of MSM factors [1],[2]. We implement this method inside a TRIPOLI4® [3] sequence in order to establish predictive mapping of MSM factors and figure out optimal position for instrumentation location at the beginning of reactor operations. Firstly, MSM factors are introduced, along with the designer point of view for geometry construction based on ROOT package [4]. At this point, the methodology of TRIPOLI4® calculation is explained in detail, including the sequencing associated to and how the Green Functions are performed within TRIPOLI4®. In this second part and within the verification framework, the previous method is extended to a “fictitious core” developed in TechnicAtome for Monte Carlo [5] calculation and for different core pattern loadings. After the completion of these numerical validations gained on a High Performing Cluster, the method is then expanded to critical mock up [6] and challenged to recent experimental results for validation. The comparisons end up with a good agreement between predictive calculation and experimental values of reactivity worth. Finally the document ends with a mid-term projection for outlooks and improvements, for ensuring an enhancement of the safety approach. Several items are discussed especially, fine tuning for the spatial meshing (regarding instrumentation size) and the impact on TRIPOLI4® Monte Carlo code with the development of new features. Then, the authors focus on sensitivity effect concerning delayed neutron spectrum and kinetics parameters. As a conclusion, this paper proposes to validate the method exposed in the near future, using experimental data coming from many years of critical mock up operations.
Highlights
Nuclear reactors operation for any purpose requires the presence of neutronic poisons to instantaneously stop the chain reaction
This paper presents a calculation route for ݂ெௌெ determination, correct rod worth measurements from boron lined proportional counter, to consider the different motion of both global and local neutron fluxes during rod drop
The methodology deals with 2 linked steps: x first, critical situation before scram is simulated for determination of both fission rate and (݊, ߙ) reaction rates distributions ; x fission rate distribution is transformed into delayed neutron source distribution for second step, considering that the delayed neutron distribution doesn’t change during scram
Summary
Nuclear reactors operation for any purpose requires the presence of neutronic poisons to instantaneously stop the chain reaction. These safety devices have several forms and the most commonly used is dropping rods made of neutron absorbing materials (Hf, AIC, B4C ...). Monte-Carlo (MC) codes can reproduce with a high fidelity the different neutron flux distributions before and after scram, and can simulate neutron propagation all the way to detector sensitivity area (boron layer for example). This paper proposes a Monte-Carlo (MC) sequence to determine the so-called MSM (Modified neutron Source Method) factors [1] [2] and to use them in order to correct the raw rod control worth measurement from Boron-lined proportional counters. Perspectives and improvements of this calculation sequence are discussed
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