Key Nuclear Data Impacting Reactivity in Advanced Reactors

Bostelmann, Friederike; Ilas, Germina; Wieselquist, William

doi:10.2172/1649145

Cited by 9 publications

(8 citation statements)

References 38 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…The large uncertainties caused by the nuclear data in HTGR simulations have been reported in some previous work (Bostelmann et al, 2020;Zhang et al, 2020). In our simulation of the HTR-10 operation history, it is found that the keff obtained by ENDF/B VII.0 is about 0.7% higher than that by ENDF/B VIII.0.…”

Section: Results and Analysissupporting

confidence: 51%

Simulation of the HTR-10 Operation History With the PANGU Code

et al. 2021

View full text Add to dashboard Cite

The 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) is the first High Temperature Gas-cooled Reactor (HTGR) in China, which was operated from January 2003 to May 2007. The HTR-10 operation history provides very important data for the validation of HTGR codes. In this paper, the HTR-10 operation history is simulated with the PANGU code, which has been recently developed for HTGR reactor physics analysis and design. Models and parameters are constructed based on the measured data of the actual conditions. The simulation results agree well with the measurements in all steady-state power periods. The discrepancy of keff is generally below 0.5%, and the discrepancy of coolant outlet temperature is generally below 5°C. It is also figured out that the burnup of graphite impurities has considerable influence on the keff at the end of the operation history, which can cause over 1.5% discrepancy when neglecting the burnup of graphite impurities. By this work, the PANGU code’s applicability in actual HTGR fuel cycle simulations is demonstrated.

show abstract

Section: Results and Analysissupporting

confidence: 51%

Simulation of the HTR-10 Operation History With the PANGU Code

et al. 2021

View full text Add to dashboard Cite

show abstract

“…Phases 1 and 2 were concluded, and the outcome has been published (Bostelmann et al, 2020a). This report summarizes the findings of Phase 3.…”

Section: Introductionmentioning

confidence: 90%

Relevant Advanced Reactor Benchmarks for Nuclear Data Assessment

Bostelmann¹,

Walker²,

Skutnik³

et al. 2020

Self Cite

View full text Add to dashboard Cite

“…The energy deposited in the core is then dissipated by conduction and can currently only be removed at the radial boundary of the PR, as depicted in Figure 8. In the gap between the PR and RPV, assumed to be 62.5-cm wide 8 , three phenomena are modeled:…”

Section: Heat Transfer Pathsmentioning

confidence: 99%

“…• Convection: During normal operations, the cold fluid flows upward in the gap. A convective 8 The PR outer and RPV inner radii are 212.5 and 275 cm, respectively boundary condition is applied to the inner side of the RPV:…”

Section: Heat Transfer Pathsmentioning

confidence: 99%

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

Labouré

Lindell

Ortensi

et al. 2022

View full text Add to dashboard Cite

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

show abstract

Key Nuclear Data Impacting Reactivity in Advanced Reactors

Cited by 9 publications

References 38 publications

Simulation of the HTR-10 Operation History With the PANGU Code

Simulation of the HTR-10 Operation History With the PANGU Code

Relevant Advanced Reactor Benchmarks for Nuclear Data Assessment

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

Contact Info

Product

Resources

About