Published 1979 | Version v1
Book

Thermal-hydraulic and structural safety analysis of SLSF P3 experiment

  • 1. Argonne National Lab., IL (USA). Reactor Analysis and Safety Div.
  • 2. Notre Dame Univ., IN (USA). Dept. of Aerospace and Mechanical Engineering

Description

The Sodium Loop Safety Facility (SLSF) P3 experiment was the fourth in a series of in-reactor tests in the Engineering Test Reactor and simulated an unprotected flow coastdown in a 37 pin bundle. A comprehensive thermal-hydraulic analysis of the SLSF loop was coupled with a structural analysis of the test section hexcan outer duct in order to provide assurance, before the transient, that the loop would safely contain the P3 experiment. This analysis was performed for both the expected transient and for an off-normal condition of failure of redundant logic circuits to provide the proper pump power demand signal. Results of the analysis showed the hexcan outer duct to safely contain the P3 experiment for both conditions. The analysis for the expected transient has been verified by the successful completion of the P3 experiment. (orig.)

Additional details

Publishing Information

Publisher
North-Holland Publishing Co.
Imprint Place
Amsterdam, Netherlands
ISBN
0444 85361 8
Imprint Title
Structural mechanics in reactor technology. Transactions. Vol. E
Journal Page Range
p. E3/5 (1-6).

Conference

Title
5. international conference on structural mechanics in reactor technology (SMIRT-5). 9. international seminar and 2. international seminar on structural reliability of mechanical components and subassemblies of nuclear power plants and 2. international seminar on containment of fast breeder reactors (CONFABRE-2).
Dates
9 - 21 Aug 1979.
Place
Berlin, Germany, F.R.

INIS

Country of Publication
Netherlands
Country of Input or Organization
Netherlands
INIS RN
12589128
Subject category
S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; S42: ENGINEERING;
Resource subtype / Literary indicator
Conference
Descriptors DEI
HYDRAULICS; LMFBR TYPE REACTORS; REACTOR SAFETY; SODIUM; THERMAL ANALYSIS
Descriptors DEC
ALKALI METALS; BREEDER REACTORS; ELEMENTS; EPITHERMAL REACTORS; FAST REACTORS; FBR TYPE REACTORS; LIQUID METAL COOLED REACTORS; METALS; REACTORS; SAFETY

Optional Information

Secondary number(s)
INKA-Conf--79-321-161.