Effects of improved modeling on best estimate BWR severe accident analysis Metadata

Title

• Main Title Effects of improved modeling on best estimate BWR severe accident analysis

Creator

• Author: Hyman, C.R.
  Creator Type: Personal
• Author: Ott, L.J.
  Creator Type: Personal

Publisher

• Name: Oak Ridge National Laboratory
  Place of Publication: Tennessee

Date

• Creation: 1984-01-01

Language

• English

Description

• Content Description: Since 1981, ORNL has completed best estimate studies analyzing several dominant BWR accident scenarios. These scenarios were identified by early Probabilistic Risk Assessment (PRA) studies and detailed ORNL analysis complements such studies. In performing these studies, ORNL has used the MARCH code extensively. ORNL investigators have identified several deficiencies in early versions of MARCH with regard to BWR modeling. Some of these deficiencies appear to have been remedied by the most recent release of the code. It is the purpose of this paper to identify several of these deficiencies. All the information presented concerns the degraded core thermal/hydraulic analysis associated with each of the ORNL studies. This includes calculations of the containment response. The period of interest is from the time of permanent core uncovery to the end of the transient. Specific objectives include the determination of the extent of core damage and timing of major events (i.e., onset of Zr/H/sub 2/O reaction, initial clad/fuel melting, loss of control blade structure, etc.). As mentioned previously the major analysis tool used thus far was derived from an early version of MARCH. BWRs have unique features which must be modeled for best estimate severe accident analysis. ORNL has developed and incorporated into its version of MARCH several improved models. These include (1) channel boxes and control blades, (2) SRV actuations, (3) vessel water level, (4) multi-node analysis of in-vessel water inventory, (5) comprehensive hydrogen and water properties package, (6) first order correction to the ideal gas law, and (7) separation of fuel and cladding. Ongoing and future modeling efforts are required. These include (1) detailed modeling for the pressure suppression pool, (2) incorporation of B/sub 4/C/steam reaction models, (3) phenomenological model of corium mass transport, and (4) advanced corium/concrete interaction modeling. 10 references, 17 figures, 1 table.
• Physical Description: 22 pages

Subject

• Keyword: Bwr Type Reactors
• Keyword: Computer Codes
• Keyword: Computer Calculations
• Keyword: Mathematical Models
• Keyword: Reactor Components
• STI Subject Categories: 22 General Studies Of Nuclear Reactors
• Keyword: Reactor Accidents
• STI Subject Categories: 21 Specific Nuclear Reactors And Associated Plants
• Keyword: Water Cooled Reactors
• Keyword: Simulation
• Keyword: Containment
• Keyword: Accidents
• Keyword: Reactors
• Keyword: Water Moderated Reactors 220900* -- Nuclear Reactor Technology-- Reactor Safety
• Keyword: M Codes
• Keyword: Forecasting
• Keyword: Computerized Simulation
• Keyword: Engineered Safety Systems
• STI Subject Categories: 210100 -- Power Reactors, Nonbreeding, Light-Water Moderated, Boiling Water Cooled
• Keyword: Reactor Cores
• Keyword: Damage
• Keyword: Containment Systems

Source

• Conference: 12. water reactor safety research information meeting, Gaithersburg, MD, USA, 23 Oct 1984

Collection

• Name: Office of Scientific & Technical Information Technical Reports
  Code: OSTI

Institution

• Name: UNT Libraries Government Documents Department
  Code: UNTGD

Resource Type

• Article

Format

• Text

Identifier

• Other: DE85000593
• Report No.: CONF-8410142-8
• Grant Number: AC05-84OR21400
• Office of Scientific & Technical Information Report Number: 6278834
• Archival Resource Key: ark:/67531/metadc1109455

Note

• Display Note: NTIS, PC A 02/MF A01; 1.