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Title: Natural convection heat transport in a small, HLMC reactor system

Abstract

Concepts are being developed and evaluated at Argonne National Laboratory for a small nuclear steam supply system (NSSS) with proliferation-resistant features targeted for export to developing countries. Here the authors are specifically investigating how simple and compact such a system can be. A heavy-liquid-metal coolant (HLMC) is being considered owing to its excellent heat transport characteristics and its relative inertness with the reference thermodynamic working fluid (water/steam). The purpose of the present work is to explore the possibility to take advantage of these HLMC characteristics by eliminating the intermediate loop needed in sodium-cooled systems and additionally eliminating the primary system coolant pumps. The criteria imposed on the system include the following: (1) low power, i.e., 300 MW(thermal); (2) small size for factory fabrication and overland transportation; (3) elimination of fuel access at the site (no refueling, fuel shuffling, nor storage at site); integral fueled module replacement at 15-yr goal interval; and (4) completion of all research and development needed for detailed prototype design within 5 yr. To accomplish the latter requirement, the authors are addressing whether existing coolant and materials technology is capable of supporting the sought-after simplifications. In this regard, they are at present considering technology developed in Russiamore » for Pb-Bi eutectic as a reactor coolant and ferritic-martensitic stainless steel with oxide-layer corrosion protection as cladding. The figure of merit in the investigation is the peak cladding temperature insofar as the cladding technology is considered proven to {approximately}600 C.« less

Authors:
; ;  [1]
  1. Argonne National Lab., IL (United States)
Publication Date:
OSTI Identifier:
678149
Report Number(s):
CONF-990605-
Journal ID: TANSAO; ISSN 0003-018X; TRN: 99:009134
Resource Type:
Journal Article
Journal Name:
Transactions of the American Nuclear Society
Additional Journal Information:
Journal Volume: 80; Conference: 1999 annual meeting of the American Nuclear Society (ANS), Boston, MA (United States), 6-10 Jun 1999; Other Information: PBD: 1999
Country of Publication:
United States
Language:
English
Subject:
21 NUCLEAR POWER REACTORS AND ASSOCIATED PLANTS; NATURAL CONVECTION; HEAT TRANSFER; DESIGN; NUCLEAR TRADE; LIQUID METAL COOLED REACTORS; PACKAGE REACTORS

Citation Formats

Spencer, B.W., Sienicki, J.J., and Farmer, M.T. Natural convection heat transport in a small, HLMC reactor system. United States: N. p., 1999. Web.
Spencer, B.W., Sienicki, J.J., & Farmer, M.T. Natural convection heat transport in a small, HLMC reactor system. United States.
Spencer, B.W., Sienicki, J.J., and Farmer, M.T. Wed . "Natural convection heat transport in a small, HLMC reactor system". United States.
@article{osti_678149,
title = {Natural convection heat transport in a small, HLMC reactor system},
author = {Spencer, B.W. and Sienicki, J.J. and Farmer, M.T.},
abstractNote = {Concepts are being developed and evaluated at Argonne National Laboratory for a small nuclear steam supply system (NSSS) with proliferation-resistant features targeted for export to developing countries. Here the authors are specifically investigating how simple and compact such a system can be. A heavy-liquid-metal coolant (HLMC) is being considered owing to its excellent heat transport characteristics and its relative inertness with the reference thermodynamic working fluid (water/steam). The purpose of the present work is to explore the possibility to take advantage of these HLMC characteristics by eliminating the intermediate loop needed in sodium-cooled systems and additionally eliminating the primary system coolant pumps. The criteria imposed on the system include the following: (1) low power, i.e., 300 MW(thermal); (2) small size for factory fabrication and overland transportation; (3) elimination of fuel access at the site (no refueling, fuel shuffling, nor storage at site); integral fueled module replacement at 15-yr goal interval; and (4) completion of all research and development needed for detailed prototype design within 5 yr. To accomplish the latter requirement, the authors are addressing whether existing coolant and materials technology is capable of supporting the sought-after simplifications. In this regard, they are at present considering technology developed in Russia for Pb-Bi eutectic as a reactor coolant and ferritic-martensitic stainless steel with oxide-layer corrosion protection as cladding. The figure of merit in the investigation is the peak cladding temperature insofar as the cladding technology is considered proven to {approximately}600 C.},
doi = {},
journal = {Transactions of the American Nuclear Society},
number = ,
volume = 80,
place = {United States},
year = {1999},
month = {9}
}