Dynamic event tree analysis with the SAS4A/SASSYS-1 safety analysis code

Jankovsky, Zachary K.; Denman, Matthew R.; Aldemir, Tunc

doi:10.1016/j.anucene.2018.01.001

Title: Dynamic event tree analysis with the SAS4A/SASSYS-1 safety analysis code

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Abstract

The consequences of a transient in an advanced sodium-cooled fast reactor are difficult to capture with the traditional approach to probabilistic risk assessment (PRA). Numerous safety-relevant systems are passive and may have operational states that cannot be represented by binary success or failure. In addition, the specific order and timing of events may be crucial which necessitates the use of dynamic PRA tools such as ADAPT. The modifications to the SAS4A/SASSYS-1 sodium-cooled fast reactor safety analysis code for linking it to ADAPT to perform a dynamic PRA are described. A test case is used to demonstrate the linking process and to illustrate the type of insights that may be gained with this process. Finally, newly-developed dynamic importance measures are used to assess the significance of reactor parameters/constituents on calculated consequences of initiating events.

Authors:

Jankovsky, Zachary K. ^[1]; Denman, Matthew R. ^[2]; Aldemir, Tunc ^[3]

The Ohio State Univ., Columbus, OH (United States). Nuclear Engineering Program; Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Risk & Reliability Analysis
Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Risk & Reliability Analysis
The Ohio State Univ., Columbus, OH (United States). Nuclear Engineering Program

Publication Date:: 2018-02-02

Research Org.:: Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

Sponsoring Org.:: USDOE Office of Nuclear Energy (NE), Fuel Cycle Technologies (NE-5)

OSTI Identifier:: 1431214

Report Number(s):: SAND-2018-2366J
Journal ID: ISSN 0306-4549; 661558; TRN: US1802644

Grant/Contract Number:: AC04-94AL85000; NA0003525

Resource Type:: Accepted Manuscript

Journal Name:: Annals of Nuclear Energy (Oxford)

Additional Journal Information:: Journal Name: Annals of Nuclear Energy (Oxford); Journal Volume: 115; Journal Issue: C; Journal ID: ISSN 0306-4549

Publisher:: Elsevier

Country of Publication:: United States

Language:: English

Subject:: 22 GENERAL STUDIES OF NUCLEAR REACTORS; ADAPT; SAS4A/SASSYS-1; SFR; DET; Dynamic PRA; Transient overpower

Citation Formats


                    Jankovsky, Zachary K., Denman, Matthew R., and Aldemir, Tunc. Dynamic event tree analysis with the SAS4A/SASSYS-1 safety analysis code.  United States: N. p., 2018. 
Web.  doi:10.1016/j.anucene.2018.01.001.

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                    Jankovsky, Zachary K., Denman, Matthew R., & Aldemir, Tunc. Dynamic event tree analysis with the SAS4A/SASSYS-1 safety analysis code.  United States.  https://doi.org/10.1016/j.anucene.2018.01.001

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                    Jankovsky, Zachary K., Denman, Matthew R., and Aldemir, Tunc. Fri .  
"Dynamic event tree analysis with the SAS4A/SASSYS-1 safety analysis code".  United States.  https://doi.org/10.1016/j.anucene.2018.01.001.  https://www.osti.gov/servlets/purl/1431214.

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@article{osti_1431214,

  title        = {Dynamic event tree analysis with the SAS4A/SASSYS-1 safety analysis code},

  author       = {Jankovsky, Zachary K. and Denman, Matthew R. and Aldemir, Tunc},

  abstractNote = {The consequences of a transient in an advanced sodium-cooled fast reactor are difficult to capture with the traditional approach to probabilistic risk assessment (PRA). Numerous safety-relevant systems are passive and may have operational states that cannot be represented by binary success or failure. In addition, the specific order and timing of events may be crucial which necessitates the use of dynamic PRA tools such as ADAPT. The modifications to the SAS4A/SASSYS-1 sodium-cooled fast reactor safety analysis code for linking it to ADAPT to perform a dynamic PRA are described. A test case is used to demonstrate the linking process and to illustrate the type of insights that may be gained with this process. Finally, newly-developed dynamic importance measures are used to assess the significance of reactor parameters/constituents on calculated consequences of initiating events.},

  doi          = {10.1016/j.anucene.2018.01.001},

  journal      = {Annals of Nuclear Energy (Oxford)},

  number       = C,

  volume       = 115,

  place        = {United States},

  year         = {2018},

  month        = {2}

}

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