Publications

Milestones Reports

State-of-the-Art Reactor Consequence Analyses (SOARCA) Project Volume 1: Peach Bottom Integrated Analysis; ADAMS Accession No. ML13150A053.

State-of-the-Art Reactor Consequence Analyses (SOARCA) Project Volume 2: Surry Integrated Analysis; ADAMS Accession No. ML13240A242.

State-of-the-Art Reactor Consequence Analyses (SOARCA) Project: Surry Uncertainty Analysis, Draft Report; ADAMS Accession No. ML15224A001.

State-of-the-Art Reactor Consequence Analyses (SOARCA) Project: Uncertainty Analysis of the Unmitigated Long-Term Station Blackout of the Peach Bottom Atomic Power Station; ADAMS Accession No. ML16133A461.

State-of-the-Art Reactor Consequence Analyses (SOARCA) Project: Sequoyah Integrated Deterministic and Uncertainty Analyses; ADAMS Accession No. ML16096A374.

State-of-the-Art Reactor Consequence Analyses (SOARCA) Project: Sequoyah Integrated Deterministic and Uncertainty Analyses; ADAMS Accession No. ML19296B786.

State-of-the-Art Reactor Consequence Analyses (SOARCA) Project: Uncertainty Analysis of the Unmitigated Short-Term Station Blackout of the Surry Power Station; ADAMS Accession.

State-of-the-Art Reactor Consequence Analysis (SOARCA) Project; ADAMS Accession No. ML12332A057.

MELCOR Best Practices as Applied in the State-of-the-Art Reactor Consequence Analyses (SOARCA) Project; https://www.nrc.gov/docs/ML1423/ML14234A136.pdf

Accident Source Terms for Light-Water Nuclear Power Plants Using High-Burnup or MOX Fuel https://www.osti.gov/servlets/purl/1010412-SeEvvB/

An Overview of the State-of-the-Art Reactor Consequence Uncertainty Assessment Accident Progression Insights https://www.osti.gov/biblio/1637263

MELCOR 2.2 Benchmarks of Peach Bottom NUREG/CR-7155 Uncertainty Analysis https://www.osti.gov/biblio/1763575

Accident Source Terms for Boiling Water Reactors with High Burnup Cores Calculated Using MELCOR 1.8.5 https://www.osti.gov/servlets/purl/1004364/

Accident Source Terms for Pressurized Water Reactors with High Burnup Cores Calculated Using MELCOR 1.8.5 https://www.osti.gov/biblio/984120

Accident Source Terms for Pressurized Water Reactors with High Burnup Cores Calculated Using MELCOR 1.8.5 https://www.osti.gov/biblio/1431254

MELCOR 2.2 Benchmarks of Peach Bottom NUREG/CR 7155 Uncertainty Analysis; https://www.osti.gov/biblio/1763575

Summary of the Uncertainty Analyses for the State-of-the-Art Reactor Consequence Analyses Project; ADAMS Accession No. ML22193A244.

MELCOR Accident Progression and Source Term Demonstration Calculations for a FHR; https://www.osti.gov/biblio/1854081

MELCOR Accident Progression and Source Term Demonstration Calculations for a Heat Pipe Reactor; https://www.osti.gov/biblio/1854082

MELCOR Accident Progression and Source Term Demonstration Calculations for HTGR;

https://www.nrc.gov/docs/ML2214/ML22144A190.pdf Fukushima Daiichi Accident Study (Status as of April 2012); https://www.osti.gov/servlets/purl/1055601

Insights Gained from Forensic Analysis with MELCOR of the Fukushima-Daiichi Accidents; https://www.osti.gov/biblio/1399212

Interim MELCOR Simulation of the Fukushima Daiichi Unit 2 Accident Reactor Core Isolation Cooling Operation; https://www.osti.gov/biblio/1325949

Modular Accident Analysis Program (MAAP)  -MELCOR Crosswalk: Phase 1 Study; https://www.epri.com/research/products/3002004449

Modular Accident Analysis Program (MAAP) – MELCOR Crosswalk: Phase II Analyzing a Partially Recovered Accident Scenario; https://www.osti.gov/biblio/1408388

Recent Papers

Beeny, B.A., et.al., “MELCOR Integrated Severe Accident Code for High Temperature Gas Cooled Reactor Applications,” NURETH-19, Brussels, Belgium, March 2022.

Wagner, K.C., et.al., “MELCOR Integrated Severe Accident Code Application To Safety Assessment Of High-Temperature Gas-Cooled Reactors,” NURETH-19, Brussels, Belgium, March 2022

Wagner, K.C., et.al., “MELCOR Integrated Severe Accident Code Application To Analysis Of Heat Pipe Reactors,” NURETH-19, Brussels, Belgium, March 2022.

David L.Y., et.al., “MELCOR Code Validation Studies on Fire Accidents in Non-Reactor Facilities,” NURETH-19, Brussels, Belgium, March 2022.

Gelbard, F., B. A. Beeny, and L. L. Humphries, “Application of MELCOR for Simulating Molten Salt Reactor Accidents,” SAND2021-15575C, NURETH-19, Brussels, Belgium, March 6-11, 2022.

Beeny, B.A., et.al., “MELCOR Integrated Severe Accident Code for Molten Salt Reactor Applications Including Fluid-Fuel Point Reactor Kinetics Equation,” NURETH-19, Brussels, Belgium, March 2022.

Schmidt, R.C., K.C. Wagner, “Modeling Heat Pipe Reactors in the MELCOR Severe Accident Code,” NURETH-19, Brussels, Belgium, March 2022.

Bixler, N.E., D.M. Osborn, J.A. Jones, C.J. Sallaberry, P.D. Mattie, and S.T. Ghosh, “SOARCA Peach Bottom Atomic Power Station Long-Term Station Blackout Uncertainty Analysis: Contributions to Overall Uncertainty,” Proceedings of the Probabilistic Safety Assessment and Management (PSAM 12) Conference, Honolulu, HI, International Association for Probabilistic Safety Assessment and Management, 2014.

Ghosh, S.T., H.E. Esmaili, A.G. Hathaway, N. Bixler, D. Brooks, D. Osborn, K. Ross, and K. Wagner, “State-of-the-Art Reactor Consequence Analyses Project Uncertainty Analyses: Insights on Methodologies,” Proceedings of the Probabilistic Safety Assessment Conference, American Nuclear Society, 2019. ADAMS Accession No. ML19064B355.

Ghosh, S.T., H.E. Esmaili, A.G. Hathaway, N. Bixler, D. Brooks, M. Dennis, D. Osborn, K. Ross, and K. Wagner, “State-of-the-Art Reactor Consequence Analyses Project: Uncertainty Analyses for Station Blackout Scenarios,” Nuclear Technology, 207(3), pp. 441–451, 2021.

Sallaberry, C.J., D.M. Osborn, N.E. Bixler, A.C. Eckert-Gallup, P.D. Mattie, and S.T. Ghosh, “SOARCA Peach Bottom Atomic Power Station Long-Term Station Blackout Uncertainty Analysis: Convergence of the Uncertainty Results,” SAND2014-1346C, Proceedings of the Probabilistic Safety Assessment and Management (PSAM 12) Conference, Honolulu, HI, International Association for Probabilistic Safety Assessment and Management, 2014.

Ghosh, S. T., et al. “State-of-the-Art Reactor Consequence Analyses Project Uncertainty Analyses: Uncertainty Analyses for Station Blackout Scenarios.” Paper 193-27145, 2019. International Topical Meeting on Probabilistic Safety Assessment and Analysis (PSA 2019), Charleston, SC, April 28-May 3, 2019.

Ghosh, S. T., et al. “State-of-the-Art Reactor Consequence Analyses Project Uncertainty Analyses: Insights on Accident Progression and Source Term.” Paper 193-27164, 2019. International Topical Meeting on Probabilistic Safety Assessment and Analysis (PSA 2019), Charleston, SC, April 28-May 3, 2019.

Ghosh, S. T., et al. “State-of-the-Art Reactor Consequence Analyses Project Uncertainty Analyses: Insights on Methodologies.” Paper 193-27166, 2019. International Topical Meeting on Probabilistic Safety Assessment and Analysis (PSA 2019), Charleston, SC, April 28-May 3, 2019. Ghosh, S. T., et al. “State-of-the-Art Reactor Consequence Analyses Project Uncertainty Analyses: Uncertainty Analyses for Station Blackout Scenarios,” Paper 193-27166, 9th Conference on Severe Accident Research (ERMSAR), Prague, Czech Republic, March 18 20, 2019.