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Title: Advanced Neutron Source Reactor (ANSR) phenomena identification and ranking (PIR) for large break loss of coolant accidents (LBLOCA)

A team of experts in reactor analysis conducted a phenomena identification and ranking (PIR) exercise for a large break loss-of-coolant accident (LBLOCA) in the Advanced Neutron source Reactor (ANSR). The LBLOCA transient is broken into two separate parts for the PIR exercise. The first part considers the initial depressurization of the system that follows the opening of the break. The second part of the transient includes long-term decay heat removal after the reactor is shut down and the system is depressurized. A PIR is developed for each part of the LBLOCA. The ranking results are reviewed to establish if models in the RELAP5-MOD3 thermalhydraulic code are adequate for use in ANSR LBLOCA simulations. Deficiencies in the RELAP5-MOD3 code are identified and existing data or models are recommended to improve the code for this application. Experiments were also suggested to establish models for situations judged to be beyond current knowledge. The applicability of the ANSR PIR results is reviewed for the entire set of transients important to the ANSR safety analysis.
Authors:
 [1] ;  [2] ;  [3] ;  [4] ;  [5]
  1. Oak Ridge National Lab., TN (United States); Tennessee Univ., Knoxville, TN (United States)
  2. Brookhaven National Lab., Upton, NY (United States)
  3. Westinghouse Savannah River Co., Aiken, SC (United States)
  4. Massachusetts Inst. of Tech., Cambridge, MA (United States)
  5. Idaho National Engineering Lab., Idaho Falls, ID (United States)
Publication Date:
OSTI Identifier:
10192861
Report Number(s):
ORNL/TM--12496
ON: DE95002440; TRN: 94:023884
DOE Contract Number:
AC05-84OR21400
Resource Type:
Technical Report
Research Org:
Oak Ridge National Lab., TN (United States)
Sponsoring Org:
USDOE, Washington, DC (United States)
Country of Publication:
United States
Language:
English
Subject:
07 ISOTOPES AND RADIATION SOURCES; 22 GENERAL STUDIES OF NUCLEAR REACTORS; NEUTRON SOURCE FACILITIES; LOSS OF COOLANT; SAFETY ANALYSIS; HEAT TRANSFER; HYDRAULICS; AFTER-HEAT REMOVAL 070201; 220900; DESIGN, FABRICATION, AND OPERATION; REACTOR SAFETY