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A diagnostic expert system for analyzing multiple-failure transients in nuclear power plants
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Source International conference on Industrial and engineering applications of artificial intelligence and expert systems archive
Proceedings of the 1st international conference on Industrial and engineering applications of artificial intelligence and expert systems - Volume 1 table of contents
Tullahoma, Tennessee, United States
Pages: 75 - 79  
Year of Publication: 1988
ISBN:0-89791-271-3
Authors
Robert P. Martin  Texas A&M Univ., College Station, TX
B. Nassersharif  Texas A&M Univ., College Station, TX
Sponsor
SIGART: ACM Special Interest Group on Artificial Intelligence
Publisher
ACM  New York, NY, USA
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ABSTRACT

CATALISP (Computer Aided Transient Analysis coded in Lisp) is a prototype expert system which is the result of a project investigating and implementing event confidence-levels (used by reactor safety experts in reactor transient analysis) in the form of an expert system. Currently, CATALISP is designed to diagnose reactor transients by analyzing simulated sensor and plant thermal hydraulic information from a system simulation. CATALISP uses a knowledge base of existing emergency nuclear plant operating guidelines and detailed thermal-hydraulic calculation results correlated to confidence-levels. CATALISP can diagnose a number of reactor transients (for example, loss-of-coolant accidents, steam-generator-tube ruptures, loss-of-offsite power, etc.). Future work includes the expansion of the knowledge base and improvement of the “deep-knowledge” qualitative models.


REFERENCES

Note: OCR errors may be found in this Reference List extracted from the full text article. ACM has opted to expose the complete List rather than only correct and linked references.

 
1
W.R. Nelson, Reactor; An Expert $.ystem for Diagnosisand Treatment of Nuclear Reactor Accidents, AAAI, August 1982.
 
2
A.H. Wells and W. E. Und erwood, "Knowledge Structures for a Nuclear Power Plant Consultant," Trans. Am. Nuc. Soc. 4_1.1, 41 (1982).
 
3
A.H. Wells and W. E. Underwood, "Reactor Operator Diagnostic Ability Training Using Knowledge-Based Computer-Aided Instruction," Trans. Am. Nuc. Soc. 46, 42 (1984).
 
4
M.A. Bray, D. E. Sebo, and B. W. Dixon, "Reactor Safety Assessment System--A Situation Assessment Aid for U SNRC Emergency Response," Expert Systems in Government Symposium, edited by Kamal N. Karna (IEEE Computer Society, 1985).
 
5
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6
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7
J.E. Suich, "Logic Programming for Operational Analysis of the Savannah River Reactors," Trans. Am. Nuc. Soc. 50, 293 (1985).
 
8
P.J. Otaduy, "Demonstration of Expert Systems in Automated Monitoring," Trans. Am. Nuc. Soc. 5_0., 298 (1985).
 
9
Knowledge Engineering Enyironmenl, Intellicorp, ver. 2.1, 1985.
 
10
Safety Code Development Group, "TRAC-PF1/MOD1, An Advanced Best-Estimate Computer Program for Pressurized Water Reactor Thermal-Hydraulic Analysis," Los Alamos National Laboratory report, LA- 10157-MS (NUREG/CR-3858).
 
11
J.F. Dearing, R. J. Henninger, and B. Nassersharif, Dominant Accident Seouences in Oconee-1 Pressurized Water Reactor, NUREG/CR-4140, LA-10351-MS, April 1985.
 
12
B. Nassersharif, Alternate St~gm ~;ene~tor Tube Rupture Mitigat~n Strategies forlhe Three Mile Island Uni~l DuringaLoss, of-Offsit~Power, LA-UR-85- 182, 1985.
 
13
N. S. DeMuth, D. Dobranich,and R. J. Henninger, Loss-of-Feedwater Transients for th~Zion- ! P~essurized WaterReactor, NUREG/CR-2656, LA- 9296-MS, May 1982.
 
14
M.A. Meyer and J. M. Booker, Sources of Correlation Between Experts' Empirical Results from Two Ex~, NUREG/CR-4814, LA- 10918-MS, April 1987.

Collaborative Colleagues:
Robert P. Martin: colleagues
B. Nassersharif: colleagues