Development of the FHR advanced natural circulation analysis code and application to FHR safety analysis

Guo, Z.; Zweibaum, N.; Shao, M.; Huddar, L. R.; Peterson, P. F.; Qiu, S.

doi:10.1016/j.pnucene.2016.03.022

Title: Development of the FHR advanced natural circulation analysis code and application to FHR safety analysis

Full Record
Other Related Research

Abstract

The University of California, Berkeley (UCB) is performing thermal hydraulics safety analysis to develop the technical basis for design and licensing of fluoride-salt-cooled, high-temperature reactors (FHRs). FHR designs investigated by UCB use natural circulation for emergency, passive decay heat removal when normal decay heat removal systems fail. The FHR advanced natural circulation analysis (FANCY) code has been developed for assessment of passive decay heat removal capability and safety analysis of these innovative system designs. The FANCY code uses a one-dimensional, semi-implicit scheme to solve for pressure-linked mass, momentum and energy conservation equations. Graph theory is used to automatically generate a staggered mesh for complicated pipe network systems. Heat structure models have been implemented for three types of boundary conditions (Dirichlet, Neumann and Robin boundary conditions). Heat structures can be composed of several layers of different materials, and are used for simulation of heat structure temperature distribution and heat transfer rate. Control models are used to simulate sequences of events or trips of safety systems. A proportional-integral controller is also used to automatically make thermal hydraulic systems reach desired steady state conditions. A point kinetics model is used to model reactor kinetics behavior with temperature reactivity feedback. The underlying large sparsemore »« less

Authors:

Guo, Z. ^[1]; Zweibaum, N. ^[2];

^[3];

^[2]; Peterson, P. F. ^[2]; Qiu, S. ^[4]

Xi'an Jiaotong Univ., Xi'an (China). Dept. of Nuclear Science and Technology; North China Electric Power Univ., Beijing (China). School of Nuclear Science and Engineering; Univ. of California, Berkeley, CA (United States). Dept. of Nuclear Engineering
Univ. of California, Berkeley, CA (United States). Dept. of Nuclear Engineering
Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States)
Xi'an Jiaotong Univ., Xi'an (China). Dept. of Nuclear Science and Technology

Publication Date:: Tue Apr 19 00:00:00 EDT 2016

Research Org.:: Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States)

Sponsoring Org.:: USDOE Office of Nuclear Energy (NE); National Natural Science Foundation of China (NSFC)

OSTI Identifier:: 1393064

Alternate Identifier(s):: OSTI ID: 1341205

Grant/Contract Number:: AC02-05CH11231; 91326201

Resource Type:: Accepted Manuscript

Journal Name:: Progress in Nuclear Energy

Additional Journal Information:: Journal Volume: 91; Journal Issue: C; Journal ID: ISSN 0149-1970

Publisher:: Elsevier

Country of Publication:: United States

Language:: English

Subject:: 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; FHR; Natural circulation; Code methodology

Citation Formats


                    Guo, Z., Zweibaum, N., Shao, M., Huddar, L. R., Peterson, P. F., and Qiu, S. Development of the FHR advanced natural circulation analysis code and application to FHR safety analysis.  United States: N. p., 2016. 
Web.  doi:10.1016/j.pnucene.2016.03.022.

Copy to clipboard


                    Guo, Z., Zweibaum, N., Shao, M., Huddar, L. R., Peterson, P. F., & Qiu, S. Development of the FHR advanced natural circulation analysis code and application to FHR safety analysis.  United States.  https://doi.org/10.1016/j.pnucene.2016.03.022

Copy to clipboard


                    Guo, Z., Zweibaum, N., Shao, M., Huddar, L. R., Peterson, P. F., and Qiu, S. Tue .  
"Development of the FHR advanced natural circulation analysis code and application to FHR safety analysis".  United States.  https://doi.org/10.1016/j.pnucene.2016.03.022.  https://www.osti.gov/servlets/purl/1393064.

Copy to clipboard


                    
@article{osti_1393064,

  title        = {Development of the FHR advanced natural circulation analysis code and application to FHR safety analysis},

  author       = {Guo, Z. and Zweibaum, N. and Shao, M. and Huddar, L. R. and Peterson, P. F. and Qiu, S.},

  abstractNote = {The University of California, Berkeley (UCB) is performing thermal hydraulics safety analysis to develop the technical basis for design and licensing of fluoride-salt-cooled, high-temperature reactors (FHRs). FHR designs investigated by UCB use natural circulation for emergency, passive decay heat removal when normal decay heat removal systems fail. The FHR advanced natural circulation analysis (FANCY) code has been developed for assessment of passive decay heat removal capability and safety analysis of these innovative system designs. The FANCY code uses a one-dimensional, semi-implicit scheme to solve for pressure-linked mass, momentum and energy conservation equations. Graph theory is used to automatically generate a staggered mesh for complicated pipe network systems. Heat structure models have been implemented for three types of boundary conditions (Dirichlet, Neumann and Robin boundary conditions). Heat structures can be composed of several layers of different materials, and are used for simulation of heat structure temperature distribution and heat transfer rate. Control models are used to simulate sequences of events or trips of safety systems. A proportional-integral controller is also used to automatically make thermal hydraulic systems reach desired steady state conditions. A point kinetics model is used to model reactor kinetics behavior with temperature reactivity feedback. The underlying large sparse linear systems in these models are efficiently solved by using direct and iterative solvers provided by the SuperLU code on high performance machines. Input interfaces are designed to increase the flexibility of simulation for complicated thermal hydraulic systems. In conclusion, this paper mainly focuses on the methodology used to develop the FANCY code, and safety analysis of the Mark 1 pebble-bed FHR under development at UCB is performed.},

  doi          = {10.1016/j.pnucene.2016.03.022},

  journal      = {Progress in Nuclear Energy},

  number       = C,

  volume       = 91,

  place        = {United States},

  year         = {Tue Apr 19 00:00:00 EDT 2016},

  month        = {Tue Apr 19 00:00:00 EDT 2016}

}

Copy to clipboard

Journal Article:

Free Publicly Available Full Text

Accepted Manuscript (Publisher)

Accepted Manuscript (DOE)

Publisher's Version of Record

https://doi.org/10.1016/j.pnucene.2016.03.022

Other availability

Search WorldCat to find libraries that may hold this journal

Citation Metrics:

Cited by: 4 works

Citation information provided by
Web of Science

Save / Share:

Export Metadata

Save to My Library

Similar Records in DOE PAGES and OSTI.GOV collections:

Sensitivity Analysis of Passive Heat Removal Capability of FHR Based on Efficient Surrogate Model and GenFlow

Journal Article Guo, Zhangpeng ; Luo, Yiping ; Niu, Fenglei - Transactions of the American Nuclear Society

Code Scaling, Applicability and Uncertainty Evaluation Methodology (CSAU) is used for quantifying reactor safety margins and works as an alternative method for licensing nuclear power plant. In the framework of CSAU, sensitivity and uncertainty analysis (SA and UA) should be performed. However, for large-scale complicated thermal hydraulics analysis simulated by code (e.g. RELAP5, CFD and subchannel analysis, the sensitivity and uncertainty analysis is really time-consuming. Furthermore, sometimes it is not feasible to perform SA and UA. Specifically, the computation time for Monte Carlo sampling or advanced LHS sampling can be neglected, but the computation time for thermal hydraulics code makesmore »« less
Integrated FHR technology development: Tritium management, materials testing, salt chemistry control, thermal hydraulics and neutronics, associated benchmarking and commercial basis

Technical Report Forsberg, Charles W. ; Peterson, Per F. ; Sridharan, Kumar ; ...

This report summarizes the results of a 3-year Integrated Research Project (IRP) sponsored by the U.S. Department of Energy to address major technical challenges in the development of the Fluoride-salt-cooled High-temperature Reactor (FHR). The IRP universities included the Massachusetts Institute of Technology (MIT), the University of California at Berkeley (UCB), the University of Wisconsin (UW), and the University of New Mexico (UNM). The FHR is a new reactor concept less than 20 years old that combines (1) a clean fluoride salt coolant, (2) the graphite-matrix coated-particle fuel originally developed for High- Temperature Gas-Cooled Reactors and (3) passive decay heat removalmore »« less
https://doi.org/10.2172/1485415

Full Text Available
MELCOR Accident Progression and Source Term Demonstration Calculations for a FHR

Technical Report Wagner, Kenneth ; Haskin, Troy ; Beeny, Brad ; ...

MELCOR is an integrated thermal hydraulics, accident progression, and source term code for reactor safety analysis that has been developed at Sandia National Laboratories for the United States Nuclear Regulatory Commission (NRC) since the early 1980s. Though MELCOR originated as a light water reactor (LWR) code, development and modernization efforts have expanded its application scope to include non-LWR reactor concepts. Current MELCOR development efforts include providing the NRC with the analytical capabilities to support regulatory readiness for licensing non-LWR technologies under Strategy 2 of the NRC's near- term Implementation Action Plans. Beginning with the Next Generation Nuclear Project (NGNP), MELCORmore »« less
https://doi.org/10.2172/1854081

Full Text Available
Impact of Passive Safety on FHR Instrumentation Systems Design and Classification

Conference Holcomb, David Eugene

Fluoride salt-cooled high-temperature reactors (FHRs) will rely more extensively on passive safety than earlier reactor classes. 10CFR50 Appendix A, General Design Criteria for Nuclear Power Plants, establishes minimum design requirements to provide reasonable assurance of adequate safety. 10CFR50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors, provides guidance on how the safety significance of systems, structures, and components (SSCs) should be reflected in their regulatory treatment. The Nuclear Energy Institute (NEI) has provided 10 CFR 50.69 SSC Categorization Guideline (NEI-00-04) that factors in probabilistic risk assessment (PRA) model insights, as well as deterministic insights, throughmore »« less
Full Text Available
Development of strong-sense validation benchmarks for the fluoride salt-cooled high-temperature reactor

Conference Blandford, E. D.

The Fluoride salt-cooled High-temperature Reactor (FHR) is a class of reactor concepts currently under development for the U. S. Dept. of Energy. The FHR is defined as a Generation IV reactor that features low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. Recent experimental work using simulant fluids have been performed to demonstrate key 'proof of principle' FHR concepts and have helped inform the reactor design process. An important element of developing FHR technology is to sufficiently validate the predictive accuracy of the computer codes used to model system response. This papermore »« less

Similar Records