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Zhang, Guanheng

Publications and source records attributed to Zhang, Guanheng.

SAM Developments to Support Transient Safety Analysis of Advanced non-LWRs

System Analysis Module (SAM) is under development at Argonne National Laboratory as a modern system-level modeling and simulation tool for advanced non-light water reactor safety analyses. It utilizes the object-oriented application framework MOOSE to leverage the modern software environment and advanced numerical methods. The capabilities of SAM are being extended to enable the transient modeling, analysis, and design of various advanced nuclear reactor systems. This report summarizes major fiscal year 2019 (FY19) progress in SAM code development, capability enhancements, demonstration, and validation to support transient safety analysis of advanced non-LWRs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Uncertainty Quantification on SAM Simulations of EBR-II Loss-of-Flow Tests

The System Analysis Module (SAM) is a modern system analysis tool being developed at Argonne National Laboratory for advanced non-LWR safety analysis. It aims to provide fast-running, whole-plant transient analyses capability with improved-fidelity for SFR, LFR, and MSR/FHR. In prior work, SAM has performed well in modeling the EBR-II benchmark based on the SHRT-17 and SHRT-45R protected and unprotected loss-of-flow experiments. However, underlying uncertainties in the modeling parameters need to be quantified to gain confidence in any best-estimate simulations. In this work, a Python coupling framework between Dakota and SAM has been developed to investigate these uncertainties, alongside code enhancements in SAM for user-flexibility in perturbing physical closure models and fluid equations of state. 25 significant input parameters covering boundary conditions, geometry, physical closure models, and equations of state were identified with their uncertainties defined from literature. Two SAM transient models for SHRT-17 and SHRT-45R were perturbed on these input parameters using Latin Hypercube Sampling and the uncertainties in predicted maximum fuel, cladding, and coolant temperatures were quantified as the output responses. Spearman rank correlation of input parameters to the predicted temperatures indicated that the most sensitive parameters include the initial core power, pipe wall drag, fuel & coolant conductivity, and peak channel flow area. Sensitivity of input parameters was found to differ between the two transients and core channel modeling strategies. The overall perturbations did not result in large uncertainties in the predicted temperatures, despite the efforts made to obtain accurate yet conservative input uncertainties.

Mui, Travis↗

Point Kinetics and Reactivity Feedback Modeling in SAM

Over the past few years, the rapidly rising interest in advanced nuclear reactor technology has led to an increased need for the developments and applications of advanced computational tools for modeling and simulation. System Analysis Module (SAM) is being developed at Argonne National Laboratory as a modern system-level modeling and simulation tool for advanced non-light water reactor safety analyses [1]. It utilizes the object-oriented application framework MOOSE [2] to leverage the modern software environment and advanced numerical methods [3]. To meet this need, the capabilities of SAM are being extended to enable the transient modeling, analysis, and design of advanced nuclear reactor systems. The objective of this summary is to present the development of Point-Kinetics modeling and reactivity feedback mechanisms in SAM. This summary also presents the new capability for modeling the thermal expansion reactivity feedback by coupling SAM and an external thermomechanics module (Tensor Mechanics) from MOOSE.

Hu, Guojun↗

Preliminary Analysis of FFTF Loss-of-Flow Tests with SAS4A/SASSYS-1

The Fast Flux Test Facility (FFTF) was the most recent Sodium-cooled Fast Reactor (SFR) built in the United States. The Passive Safety Test (PST) program was carried out on FFTF between 1984 and 1986 to demonstrate the capability of passive reactor shutdown and decay heat removal in response to protected and unprotected transients. In order to recover the information and knowledge that has been gained from FFTF, a research project was initiated focusing on the simulation of Unprotected Loss of Flow Without Scram (LOFWOS) tests similar to the knowledge preservation activities conducted for EBR-II by Argonne National Laboratory in earlier years. These knowledge preservation activities are intended to support the validation of computational tools and methods in the area of SFR passive safety. Comparisons with experimental data and other safety codes provide opportunities to improve SFR design and safety analysis codes and methods.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Self-Sustaining Thorium Boiling Water Reactors

The primary objectives of this project are to: Perform a pre-conceptual design of a core for an alternative to the Hitachi proposed fuel-self- sustaining RBWR-AC, to be referred to as a RBWR-Th. The use of thorium fuel is expected to assure negative void coefficient of reactivity (versus positive of the RBWR-AC) and improve reactor safety; Perform a pre-conceptual design of an alternative core to the Hitachi proposed LWR TRU transmuting RBWR-TB2, to be referred to as the RBWR-TR. In addition to improved safety, use of thorium for the fertile fuel is expected to improve the TRU transmutation effectiveness; Compare the RBWR-Th and RBWR-TR performance against that of the Hitachi RBWR core designs and sodium cooled fast reactor counterparts - the ARR and ABR; and, Perform a viability assessment of the thorium-based RBWR design concepts to be identified along with their associated fuel cycle, a technology gap analysis, and a technology development roadmap. A description of the work performed and of the results obtained is provided in this Overview Report and, in more detail, in the Attachments. The major findings of the study are summarized.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗