Spontaneous stabilization of HTGRs without reactor scram and core cooling—Safety demonstration tests using the HTTR: Loss of reactivity control and core cooling

Takamatsu, Kuniyoshi; Yan, Xing; Nakagawa, Shigeaki; Sakaba, Nariaki; Kunitomi, Kazuhiko

doi:10.1016/j.nucengdes.2013.12.005

Cited by 16 publications

(14 citation statements)

References 4 publications

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“…The cold helium flows upward between the reactor pressure vessel (RPV) and permanent reflector (PR), then downward in the cooling channels inside the core. The vessel cooling system (VCS) constitutes the ultimate heat sink during loss-of-cooling conditions, with the heat from the RPV being rejected primarily through radiation and natural circulation [39]. Startup neutron sources are present in the reactor and also affect its behavior.…”

Section: Reactor Overviewmentioning

confidence: 99%

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FY22 Status Report on the ART-GCR CMVB and CNWG International Collaborations

Labouré

Lindell

Ortensi

et al. 2022

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This work presents the numerical model for the high temperature test reactor (HTTR) loss of forced cooling (LOFC) experiment with the INL codes Griffin, BISON, and RELAP-7, based on the Multiphysics Object-Oriented Simulation Environment (MOOSE) framework. Promising results were obtained, with the overall behavior of the reactor successfully captured. Changes in the heat transfer coupling, as compared to the fiscal year (FY)-21 model, enabled drastic improvement of the steady-state solution, both in terms of computational time and global energy discrepancies. The former was reduced by a factor of roughly 60, whereas the latter decreased from 11% to less than 2%. Furthermore, the discrepancy in the steady-state multiplication factor was improved from +2,300 and +2,900 pcm (for the 30 and 9 MW cases) to -700 and +1,200 pcm, respectively, and now falls well within the large measurement uncertainties stemming from graphite impurities. Validation of the Monte Carlo model used to generate cross sections was also performed against available measurements. Though significant, the discrepancies remain acceptable overall in light of the large uncertainty.Numerous improvements are still needed to better compare with the experiment involving the 9 MW case and to instill greater confidence in the model's ability to accurately predict the 30 MW behavior. Specifically, the power levels predicted by the 9 MW transient simulation following re-criticality remain low, pointing to an underestimation of the passive cooling of the core. A key aspect of future work will be to better understand the flow pattern during the LOFC event, particularly to determine if natural or forced convection is occurring inside the reactor pressure vessel (RPV). More generally, additional validation data would be immensely useful for further enhancing the numerical model and better matching the experiments. In addition, a more sophisticated thermal-hydraulics model that simulates all the channels as a single system model should be considered to take into account the rest of the primary loop. Finally, even if the results are iv in better agreement with the experiments, sensitivity analysis and uncertainty quantification will be necessary to evaluate the model uncertainty. v

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Section: Reactor Overviewmentioning

confidence: 99%

“…Table 2 provides further details on the fuel specifications. A more exhaustive description of the test reactor can be found in [6,39], among other places.…”

Section: Reactor Overviewmentioning

confidence: 99%

FY22 Status Report on the ART-GCR CMVB and CNWG International Collaborations

Labouré

Lindell

Ortensi

et al. 2022

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“…This would then allow to repeat the depletion calculation using a more accurate temperature distribution. Although it is likely that this local burnup may also be of significant importance, since the LOFC transient and the time to recriticality are highly dependent on the initial xenon concentrations, which in turn depend on the local flux level [19], the cross sections are not considered to depend thereon in this report (mostly due to time constraint). The approximated temperature profile as well as the tabulation points are also described Section 3.…”

Section: Introductionmentioning

confidence: 99%

HTTR 3 D Cross Section Generation with Serpent and MAMMOTH

Labouré

Ortensi

Hummel

2018

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The High Temperature Test Reactor (HTTR) is a graphite-moderated thermal reactor with a nominal power output of 30 MWth. The purpose of this work is to prepare a set of neutron cross sections for use in the transient analyses of three Loss of Forced Cooling (LOFC) events planned for this reactor. The cross sections are generated with the Serpent 2 3-D Monte Carlo code to include the axial heterogeneity and the strong axial coupling of this core. The approach yields good agreement of results when compared to previously used 2-D lattice models, which require a fine spatial discretization and, surprisingly, comes at a higher computational cost. In addition, the temperature distribution must be considered to obtain reasonable cross sections and the axial leakage component dominates the spectral effects in the core. Consequently, a temperature model as a function of the fuel and moderator temperatures is adopted, and the core is depleted to 390 effective full-power days (EFPDs). A full tabulation of cross sections is prepared at this burnup point for a variety of fuel and moderator temperatures. Next, the multiphysics reactor application MAMMOTH is used to evaluate the quality of these cross sections using a diffusion solver. As expected, the homogenization error in the cross sections is significant and the Super Homogenization (SPH) correction from MAMMOTH is necessary to preserve key reaction rates. The SPH-corrected MAMMOTH results are in excellent agreement with the Serpent results and reproduce both the reference power profile and temperature coefficients at each tabulation point. This fact confirms the accuracy of the SPH-corrected cross sections and demonstrates the maturity of using Serpent 2 and MAMMOTH for 3-D cross-section generation, even for reactors as complex as the HTTR. Finally, the authors recommend that the cross sections should be parametrized with the local burnup and that a full core burnup calculation with coupled thermal-fluids would provide a better estimate of the initial condition for any subsequent transient analysis. Unfortunately, this was not within the scope of this work, and direct analysis with these tabulations should be considered as a first-order approximation, at best.

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“…The major specifications of the HTTR are shown in Table 1. Safety demonstration tests were conducted to demonstrate the inherent safety features of the HTTR [1]. The central control rod (C-CR) withdrawal test was one of the HTTR safety demonstration tests besides coolant circulator trips.…”

Section: Introductionmentioning

confidence: 99%

Determination of Reactivity and Neutron Flux Using Modified Neural Network for HTGR

et al. 2017

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Nuclear kinetic calculations based on point kinetic model have been generally applied as the standard method for neutronics codes. As the central control rod (C-CR) withdrawal test has demonstrated in a prismatic core type high-temperature gas-cooled reactor (HTGR) named High Temperature Engineering Test Reactor (HTTR), the transient calculation of kinetic parameter, reactivity, and neutron fluxes, requires a new method to shorten calculation-process time. Development of neural network method was applied to point kinetic model as the necessity of real-time calculation that could work in parallel with the digital reactivity meter. The combination of Time Delayed Neural Network (TDNN) and Jordan Recurrent Neural Network (Jordan RNN) named TD-Jordan RNN was the result of the modeling approach. The application of TD-Jordan RNN with adequate learning, tested offline, determined results accurately even when signal inputs were noisy. Furthermore, the preprocessing for neural network input utilized noise reduction as one of the equations to transform two of twelve time-delayed inputs into power corrected inputs.

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Spontaneous stabilization of HTGRs without reactor scram and core cooling—Safety demonstration tests using the HTTR: Loss of reactivity control and core cooling

Cited by 16 publications

References 4 publications

FY22 Status Report on the ART-GCR CMVB and CNWG International Collaborations

FY22 Status Report on the ART-GCR CMVB and CNWG International Collaborations

HTTR 3 D Cross Section Generation with Serpent and MAMMOTH

Determination of Reactivity and Neutron Flux Using Modified Neural Network for HTGR

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